Nanocellulose-stabilized Pickering emulsions and their applications.

Science.gov (United States)

Fujisawa, Shuji; Togawa, Eiji; Kuroda, Katsushi

2017-01-01

Pickering emulsion, which is an emulsion stabilized by solid particles, offers a wide range of potential applications because it generally provides a more stable system than surfactant-stabilized emulsion. Among various solid stabilizers, nanocellulose may open up new opportunities for future Pickering emulsions owing to its unique nanosizes, amphiphilicity, and other favorable properties (e.g. chemical stability, biodegradability, biocompatibility, and renewability). In this review, the preparation and properties of nanocellulose-stabilized Pickering emulsions are summarized. We also provide future perspectives on their applications, such as drug delivery, food, and composite materials.

AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1994

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-06-01

The Pickering Nuclear Generating Station (PNGS) is located on the north shore of Lake Ontario, about 32 km east of downtown Toronto. It consists of two stations, PNGS-A and PNGS-B. Each station contains four reactor units. PNGS-A consists of Units 1 to 4, while PNGS-B consists of Units 5 to 8. Each unit can generate about 540 megawatts of electricity. All eight units are located within a single enclosure. Ontario Hydro`s Pickering Nuclear Division has assigned one Station Director with authority over both stations, but each station has its own organization. AECB issue a separate operating licence for each station. This report presents the Atomic Energy Control Board staff assessment of the Pickering stations` safety performance in 1994 and other aspects that they consider to have significant impact on nuclear safety. AECB based their conclusions on their observations, audits, inspections and review of information that Ontario Hydro submits to them as required by the station Operating Licences. 11 tabs., 8 figs.

AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1994

International Nuclear Information System (INIS)

1995-06-01

The Pickering Nuclear Generating Station (PNGS) is located on the north shore of Lake Ontario, about 32 km east of downtown Toronto. It consists of two stations, PNGS-A and PNGS-B. Each station contains four reactor units. PNGS-A consists of Units 1 to 4, while PNGS-B consists of Units 5 to 8. Each unit can generate about 540 megawatts of electricity. All eight units are located within a single enclosure. Ontario Hydro's Pickering Nuclear Division has assigned one Station Director with authority over both stations, but each station has its own organization. AECB issue a separate operating licence for each station. This report presents the Atomic Energy Control Board staff assessment of the Pickering stations' safety performance in 1994 and other aspects that they consider to have significant impact on nuclear safety. AECB based their conclusions on their observations, audits, inspections and review of information that Ontario Hydro submits to them as required by the station Operating Licences. 11 tabs., 8 figs

Commissioning quality assurance at Pickering NGS

International Nuclear Information System (INIS)

Wieckowski, J.T.

1983-05-01

Ontario Hydro decided in 1978 to implement a formal quality assurance program applicable to commissioning and operation of nuclear generating stations. Pickering NGS is the first station to have the commissioning quality assurance (CQA) program applied to it. This paper outlines the scope, implementation, and evaluation of the CQA program as applied to Pickering Unit 5

Pickering tool management system

International Nuclear Information System (INIS)

Wong, E.H.; Green, A.H.

1997-01-01

Tools were being deployed in the station with no process in effect to ensure that they are maintained in good repair so as to effectively support the performance of Maintenance activities. Today's legal requirements require that all employers have a process in place to ensure that tools are maintained in a safe condition. This is specified in the Ontario Health and Safety Act. The Pickering Tool Management System has been chosen as the process at Pickering N.D to manage tools. Tools are identified by number etching and bar codes. The system is a Windows application installed on several file servers

Seismic analysis of mechanical systems at Pickering NGS

International Nuclear Information System (INIS)

Ghobarah, A.

1995-11-01

The objective of this study is to assess the seismic withstand capacity of selected safety-related mechanical systems associated with the Pressure Relief Duct (PRD) at the Pickering A Nuclear Generating Station. These systems are attached to the PRD and include the Emergency Coolant Injection System piping, the Vacuum Ducts, the Emergency Water Storage System, the PRD expansion joint seals and the PRD to Reactor Building joint seals. The input support motion to the mechanical systems is taken to be the seismic response of the PRD determined in an earlier study using various levels of predetermined ground response spectrum envelope. (author). 12 refs., 13 tabs., 48 figs

Retrofit of AECL CAN6 seals into the Pickering shutdown cooling pumps

International Nuclear Information System (INIS)

Rhodes, D.; Metcalfe, R.; Brown, G.

1997-01-01

The existing mechanical seals in the shutdown cooling (SDC) pumps at the eight-unit Pickering Nuclear Generating Station have caused as least seven forced outages in the last fifteen years. The SDC pumps were originally intended to run only during shutdowns, mostly at low pressure, except for short periods during routine testing of SDC isolation valves while the plant is operating at full pressure to verify that the emergency core injection system is available. Unfortunately, in practice, some SDC pumps must be run much more frequently than this to prevent overheating or freezing of components in the system while the plant is at power. This more severe service has decreased seal lifetime from about 8000 running hours to about 3000 running hours. Rather than tackling the difficult task of eliminating on-power running of the pumps, Pickering decided to install a more robust seal design that could withstand this. Through the process of competitive tender, AECL's CAN6 seal was chosen. This seal has a successful history in similarly demanding conditions in boiling water reactors in the USA. To supplement this and demonstrate there would be no 'surprises,' a 2000-hour test program was conducted. Testing consisted of simulating all the expected conditions, plus some special tests under abnormal conditions. This has given assurance that the seal will operate reliably in the Pickering shutdown cooling pumps. (author)

Retrofit of AECL CAN6 seals into the Pickering shutdown cooling pumps

International Nuclear Information System (INIS)

Rhodes, D.; Metcalfe, R.; Brown, G.; Kiameh, P.; Burchett, P.

1997-01-01

The existing mechanical seals in the shutdown cooling (SDC) pumps at the eight-unit Pickering Nuclear Generating Station have caused at least seven forced outages in the last fifteen years. The SDC pumps were originally intended to run only during shutdowns, mostly at low pressure, except for short periods during routine testing of SDC isolation valves while the plant is operating at full pressure to verify that the emergency core injection system is available. Unfortunately, in practice, some SDC pumps must be run much more frequently than this to prevent overheating or freezing of components in the system while the plant is at power. This more severe service has decreased seal lifetime from about 8000 running hours to about 3000 running hours. Rather than tackling the difficult task of eliminating on-power running of the pumps, Pickering decided to install a more robust seal design that could withstand this. Through the process of competitive tender, AECL's CAN6 seal was chosen. This seal has a successful history in similarly demanding conditions in boiling water reactors in the USA. To supplement this and demonstrate there would be no 'surprises,' a 2000-hour test program was conducted. Testing consisted of simulating all the expected conditions, plus some special tests under abnormal conditions. This has given assurance that the seal will operate reliably in the Pickering shutdown cooling pumps. (author)

Bruce used fuel dry storage project evolution from Pickering to Bruce

International Nuclear Information System (INIS)

Young, R.E.

1996-01-01

Additional fuel storage capacity is required at Bruce Nuclear Generating Station, which otherwise would soon fill up all its pool storage capacity. The recommended option was to use a dry storage container similar to that at Pickering. The changes made to the Pickering type of container included: fuel to be stored in trays; the container's capacity increased to 600 bundles; the container's lid to be changed to a metal one; the single concrete lid to be changed to a double metal lid system; the container not to be transportable; the container would be dry-loaded. 7 figs

Bruce used fuel dry storage project evolution from Pickering to Bruce

Energy Technology Data Exchange (ETDEWEB)

Young, R E [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

1997-12-31

Additional fuel storage capacity is required at Bruce Nuclear Generating Station, which otherwise would soon fill up all its pool storage capacity. The recommended option was to use a dry storage container similar to that at Pickering. The changes made to the Pickering type of container included: fuel to be stored in trays; the container`s capacity increased to 600 bundles; the container`s lid to be changed to a metal one; the single concrete lid to be changed to a double metal lid system; the container not to be transportable; the container would be dry-loaded. 7 figs.

Pickering NGS A reactor building 1 dome refurbishment long-term monitoring of coating

International Nuclear Information System (INIS)

Deans, J.J.; Chan, P.; Gomme, R.

2006-01-01

'Full text:' To reduce air leakage through the dome of Pickering NGS A Reactor Building 1, in August 1993 a portion of the exterior concrete surface was coated with a single component elastomeric polyurethane material. An internal positive pressure test of the building, conducted between November 5 and 7, 1993, found that the air leakage rates were significantly lower in this test than leakage rates which had been measured during a pressure test conducted in 1992. This reduction in leakage was attributed to the successful performance of the coating. The need for a high-performance, elastomeric surface coating was identified for reduction of air leakage levels through the dome of Reactor Building l of Ontario Power Generation's (formerly Ontario Hydro's) Pickering 'A' Nuclear Generating Station near Toronto. A number of candidate coatings were extensively tested to assess the performance characteristics and identify a material that could withstand the elements and perform effectively for around 20 years. Under normal operating conditions, a licensing limit of 2.7% of contained mass/hour is set for permissible containment leakage whilst the operational working target is less than 1%. The facility's engineers determined that any leakages were pressure-dependent, so in an effort to remain well within their working target, they sought a system that would bridge and seal any hairline cracks in the concrete dome and thereby prevent the passage of gas or vapour through the substrate. On the basis of scheduling and cost, they concluded that a high performance coating was most appropriate for the project, and hired Kinectrics (formerly Ontario Hydro Technologies (OHT)) to select, test, assess and arrange for the application to the RB 1 Dome. In all, nearly 70 separate manufacturers were approached by Kinectrics with a view to obtaining recommendations for treatment. The respective performance data of the respondents' products were compared with a set of specific design

Analytical Diagnostics of Non-Optimal Use of Pesticides and Health Hazards for Vegetable Pickers

International Nuclear Information System (INIS)

Zafar, M.; Mehmood, T.; Baig, I. A.; Saboor, A.; Sadiq, S.; Mahmood, K.

2016-01-01

Economically pesticides are meant to control pests in the fields. Up to certain optimal use of a typical pesticide, it enhances the yield of crops and vegetables. But, eventually amplified use of pesticides results in contamination of environment (water, soil, and air) and increase the health cost of vegetable pickers. The purpose of this study is to estimate the excessive use of pesticides and economic cost of health hazards for the vegetable pickers in district Vehari. Data from 90 respondents were collected and analyzed. The most common health problems identified during the survey were headache, eye irritation, skin infection, cough and shortness of breath. Health cost consists of costs related to precautionary measure, medication, traveling, the opportunity cost of attended persons and productivity loss. The mean health cost of vegetable pickers in the study area was about Rs. 385 per picker per year. Health cost model was used to measure the health cost of vegetable pickers. The regression results showed that pesticides were being applied non-optimally in the study area i.e., number of pesticide applications for vegetables (7-31) were substantially higher than the recommended dose. Health cost function was significantly different from zero as indicated by F-stat (32.18) and it is also supported by R/sup 2/ that about 70 percent variation in health cost is explained by medication accompanied by productivity loss (Rs. 223), precautionary measure (Rs. 134), attended person cost (Rs. 14) and traveling expenditures (Rs. 16). Hence, strict legislation is required to overcome the availability of hazardous pesticides and to keep the vegetable pickers aware of the optimal use of pesticides through appropriate extension services. (author)

Estimating the possible range of recycling rates achieved by dump waste pickers: The case of Bantar Gebang in Indonesia.

Science.gov (United States)

Sasaki, Shunsuke; Araki, Tetsuya

2014-06-01

This article presents informal recycling contributions made by scavengers in the surrounding area of Bantar Gebang final disposal site for municipal solid waste generated in Jakarta. Preliminary fieldwork was conducted through daily conversations with scavengers to identify recycling actors at the site, and then quantitative field surveys were conducted twice. The first survey (n = 504 households) covered 33% of all households in the area, and the second survey (n = 69 households) was conducted to quantify transactions of recyclables among scavengers. Mathematical equations were formulated with assumptions made to estimate the possible range of recycling rates achieved by dump waste pickers. Slightly over 60% of all respondents were involved in informal recycling and over 80% of heads of households were waste pickers, normally referred to as live-in waste pickers and live-out waste pickers at the site. The largest percentage of their spouses were family workers, followed by waste pickers and housewives. Over 95% of all households of respondents had at least one waste picker or one small boss who has a coequal status of a waste picker. Average weight of recyclables collected by waste pickers at the site was estimated to be approximately 100 kg day(-1) per household on the net weight basis. The recycling rate of solid wastes collected by all scavengers at the site was estimated to be in the range of 2.8-7.5% of all solid wastes transported to the site. © The Author(s) 2014.

Pickering education centre aids nuclear acceptance

International Nuclear Information System (INIS)

Anon.

1979-01-01

Activities at the new education centre at Pickering are described. The opening of the Nuclear Communications Centre, in 1978, resulting from a search for an effective means of maintaining public acceptance of Ontario Hydro's extensive nuclear power programme. Activities include participation in the interactive computer games, guided tours of educational exhibits including a model of Pickering A generating station, and displays depicting the Candu fuel cycle, outdoor exhibits of renewable energy sources, and tours of the plant. Outside activities include lectures to schools and citizen, business, or professional groups. (U.K.)

Data-driven warehouse optimization : Deploying skills of order pickers

NARCIS (Netherlands)

M. Matusiak (Marek); M.B.M. de Koster (René); J. Saarinen (Jari)

2015-01-01

textabstractBatching orders and routing order pickers is a commonly studied problem in many picker-to-parts warehouses. The impact of individual differences in picking skills on performance has received little attention. In this paper, we show that taking into account differences in the skills of

Estimating cancer risk in relation to tritium exposure from routine operation of a nuclear-generating station in Pickering, Ontario.

Science.gov (United States)

Wanigaratne, S; Holowaty, E; Jiang, H; Norwood, T A; Pietrusiak, M A; Brown, P

2013-09-01

Evidence suggests that current levels of tritium emissions from CANDU reactors in Canada are not related to adverse health effects. However, these studies lack tritium-specific dose data and have small numbers of cases. The purpose of our study was to determine whether tritium emitted from a nuclear-generating station during routine operation is associated with risk of cancer in Pickering, Ontario. A retrospective cohort was formed through linkage of Pickering and north Oshawa residents (1985) to incident cancer cases (1985-2005). We examined all sites combined, leukemia, lung, thyroid and childhood cancers (6-19 years) for males and females as well as female breast cancer. Tritium estimates were based on an atmospheric dispersion model, incorporating characteristics of annual tritium emissions and meteorology. Tritium concentration estimates were assigned to each cohort member based on exact location of residence. Person-years analysis was used to determine whether observed cancer cases were higher than expected. Cox proportional hazards regression was used to determine whether tritium was associated with radiation-sensitive cancers in Pickering. Person-years analysis showed female childhood cancer cases to be significantly higher than expected (standardized incidence ratio [SIR] = 1.99, 95% confidence interval [CI]: 1.08-3.38). The issue of multiple comparisons is the most likely explanation for this finding. Cox models revealed that female lung cancer was significantly higher in Pickering versus north Oshawa (HR = 2.34, 95% CI: 1.23-4.46) and that tritium was not associated with increased risk. The improved methodology used in this study adds to our understanding of cancer risks associated with low-dose tritium exposure. Tritium estimates were not associated with increased risk of radiationsensitive cancers in Pickering.

From pioneering to implementing automated blood pressure measurement in clinical practice: Thomas Pickering's legacy

DEFF Research Database (Denmark)

Stolarz-Skrzypek, Katarzyna; Thijs, Lutgarde; Wizner, Barbara

2010-01-01

Thomas G. Pickering spent most of his scientific career in carrying out research on clinical hypertension and blood pressure (BP) measurement. In our review of Pickering's seminal work, we first focused on white-coat hypertension and masked hypertension, two terms that he had introduced. Next, we...... highlighted the early publications of Pickering on diurnal BP variability and on the clinical application of self-measured BP. Pickering's work inspired many investigators worldwide and constituted a solid basis for further research. Pickering's original ideas led to algorithms for risk stratification...

Pickering safeguards: a preliminary analysis

International Nuclear Information System (INIS)

Todd, J.L.; Hodgkinson, J.G.

1977-05-01

A summary is presented of thoughts relative to a systems approach for implementing international safeguards. Included is a preliminary analysis of the Pickering Generating Station followed by a suggested safeguards system for the facility

Leptospira Exposure and Waste Pickers: A Case-Control Seroprevalence Study in Durango, Mexico

Science.gov (United States)

Alvarado-Esquivel, Cosme; Hernandez-Tinoco, Jesus; Sanchez-Anguiano, Luis Francisco; Ramos-Nevarez, Agar; Cerrillo-Soto, Sandra Margarita; Guido-Arreola, Carlos Alberto

2015-01-01

Background Infection with Leptospira may occur by contact with Leptospira-infected animals. Waste pickers are in contact with rodents and dogs while picking in the garbage. Whether waste pickers are at risk for Leptospira infection is largely unknown. This study was aimed to determine the association of Leptospira IgG seroprevalence with the occupation of waste picking, and to determine the epidemiological characteristics of the waste pickers with Leptospira exposure. Methods Through a case-control study, we determined the seroprevalence of anti-Leptospira IgG antibodies in 90 waste pickers and 90 age- and gender-matched control subjects in Durango City, Mexico using an enzyme immunoassay. Data were analyzed by bivariate and multivariate analyses. Results The prevalence of anti-Leptospira IgG antibodies was similar in waste pickers (4/90: 4.4%) to that in control subjects (5/90: 5.6%) (P = 1.00). Bivariate analysis showed that Leptospira exposure in waste pickers was associated with increasing age (P = 0.009), no education (P = 0.008), and consumption of rat meat (P = 0.04). However, these associations were no longer found by multivariate analysis. Leptospira exposure in waste pickers was not associated with health status, duration in the activity, wearing hand gloves and facemasks, history of injuries with sharp material of the garbage, or contact with animals or soil. Conclusions This is the first study about Leptospira exposure in waste pickers. Results suggest that waste pickers are not at increasing risk for Leptospira exposure in Durango City, Mexico. Further research with a larger sample size to elucidate the association of Leptospira exposure with waste picking activity is needed. PMID:26124911

Leptospira Exposure and Waste Pickers: A Case-Control Seroprevalence Study in Durango, Mexico.

Science.gov (United States)

Alvarado-Esquivel, Cosme; Hernandez-Tinoco, Jesus; Sanchez-Anguiano, Luis Francisco; Ramos-Nevarez, Agar; Cerrillo-Soto, Sandra Margarita; Guido-Arreola, Carlos Alberto

2015-08-01

Infection with Leptospira may occur by contact with Leptospira-infected animals. Waste pickers are in contact with rodents and dogs while picking in the garbage. Whether waste pickers are at risk for Leptospira infection is largely unknown. This study was aimed to determine the association of Leptospira IgG seroprevalence with the occupation of waste picking, and to determine the epidemiological characteristics of the waste pickers with Leptospira exposure. Through a case-control study, we determined the seroprevalence of anti-Leptospira IgG antibodies in 90 waste pickers and 90 age- and gender-matched control subjects in Durango City, Mexico using an enzyme immunoassay. Data were analyzed by bivariate and multivariate analyses. The prevalence of anti-Leptospira IgG antibodies was similar in waste pickers (4/90: 4.4%) to that in control subjects (5/90: 5.6%) (P = 1.00). Bivariate analysis showed that Leptospira exposure in waste pickers was associated with increasing age (P = 0.009), no education (P = 0.008), and consumption of rat meat (P = 0.04). However, these associations were no longer found by multivariate analysis. Leptospira exposure in waste pickers was not associated with health status, duration in the activity, wearing hand gloves and facemasks, history of injuries with sharp material of the garbage, or contact with animals or soil. This is the first study about Leptospira exposure in waste pickers. Results suggest that waste pickers are not at increasing risk for Leptospira exposure in Durango City, Mexico. Further research with a larger sample size to elucidate the association of Leptospira exposure with waste picking activity is needed.

PROFILE OF PLASTIC WATER BOTTLES WASTES PROCESSING BUSINESS UNIT FOR WASTE PICKERS

Directory of Open Access Journals (Sweden)

Herijanto P.

2017-09-01

Full Text Available Used plastic water bottles waste pickers can be categorized as one of the informal sector’s component. They work for themselves by picking up used water bottles and selling them to the waste collectors. The problem to be solved in this research is How the Most Appropriate Used Plastic Water Bottles Business Model for Waste Pickers Is that enables them to be categorized as formal sector. From the result of the interview with 120 waste pickers, 96 results were qualified to be analyzed. The interview was located in several waste collectors, which were visited by waste pickers at certain hours. The data were analyzed descriptively based on six business aspects. Specifically for production facilities, Quality Function Deployment (QFD and Value Engineering (VE analysis were performed. The results of the analysis indicate that the business is practicable for waste pickers and has the potential to enable them run a formal business sector.

Calculation of homogenized Pickering NGS stainless steel adjuster rod neutron cross-sections using conservation of reaction rates

Energy Technology Data Exchange (ETDEWEB)

Robinson, R C [Atlantic Nuclear Services Ltd. (Canada); Tran, F [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station

1996-12-31

A homogenization methodology for calculation of reactivity device incremental cross-sections has been developed using reaction rate conservation (RRC). A heterogeneous transport calculation of flux was utilised to produce the homogenized cross-sections for a finite difference two group diffusion code. The RRC cross-sections have been shown to improve significantly the prediction of reactivity worth for stainless steel adjuster rods installed in Pickering NGS reactors. (author). 10 refs., 3 tabs., 6 figs.

Water-in-oil Pickering emulsions stabilized by stearoylated microcrystalline cellulose.

Science.gov (United States)

Pang, Bo; Liu, Huan; Liu, Peiwen; Peng, Xinwen; Zhang, Kai

2018-03-01

Hydrophobic particles with static water contact angles larger than 90° are more like to stabilize W/O Pickering emulsions. In particular, high internal phase Pickering emulsions (HIPEs) are of great interest for diverse applications. However, W/O HIPEs have rarely been realized using sustainable biopolymers. Herein, we used stearoylated microcrystalline cellulose (SMCC) to stabilize W/O Pickering emulsions and especially, W/O HIPEs. Moreover, these W/O HIPEs can be further used as platforms for the preparation of porous materials, such as porous foams. Stearoylated microcrystalline cellulose (SMCC) was prepared by modifying MCC with stearoyl chloride under heterogeneous conditions. Using SMCC as emulsifiers, W/O medium and high internal phase Pickering emulsions (MIPEs and HIPEs) with various organic solvents as continuous phases were prepared using one-step and two-step methods, respectively. Polystyrene (PS) foams were prepared after polymerization of oil phase using HIPEs as templates and their oil/water separation capacity were studied. SMCC could efficiently stabilize W/O Pickering emulsions and HIPEs could only be prepared via the two-step method. The internal phase volume fraction of the SMCC-stabilized HIPEs reached as high as 89%. Diverse internal phase volume fractions led to distinct inner structures of foams with closed or open cells. These macroporous polystyrene (PS) foams demonstrated great potential for the effective absorption of organic solvents from underwater. Copyright © 2017 Elsevier Inc. All rights reserved.

Pickering seismic safety margin

International Nuclear Information System (INIS)

Ghobarah, A.; Heidebrecht, A.C.; Tso, W.K.

1992-06-01

A study was conducted to recommend a methodology for the seismic safety margin review of existing Canadian CANDU nuclear generating stations such as Pickering A. The purpose of the seismic safety margin review is to determine whether the nuclear plant has sufficient seismic safety margin over its design basis to assure plant safety. In this review process, it is possible to identify the weak links which might limit the seismic performance of critical structures, systems and components. The proposed methodology is a modification the EPRI (Electric Power Research Institute) approach. The methodology includes: the characterization of the site margin earthquake, the definition of the performance criteria for the elements of a success path, and the determination of the seismic withstand capacity. It is proposed that the margin earthquake be established on the basis of using historical records and the regional seismo-tectonic and site specific evaluations. The ability of the components and systems to withstand the margin earthquake is determined by database comparisons, inspection, analysis or testing. An implementation plan for the application of the methodology to the Pickering A NGS is prepared

Evaluation of organic coatings to reduce air leakage through cracks in the Pickering NGS 'A' reactor building 1

International Nuclear Information System (INIS)

Deans, J.J.; Sato, J.A.; Hampton, J.H.D.; Cullen, R.; Paterson, G.; Chan, P.; Rajagopalan, R.

1994-01-01

Pressure tests conducted in 1992 on the Pickering NGS 'A' Reactor Building 1 showed that the containment leakage rate of the building was close to the licensing limit. The leakage was found to be pressure dependent and was attributed to cracks in the concrete dome. A number of solutions were studied by a task group, and the application of an organic coating to the exterior surface of the dome was identified as the most viable solution under the constraints of schedule and cost. In addition to reducing the air leakage rate, the coating material must be flexible to bridge existing moving cracks, it must have excellent adhesion to the concrete substrate to sustain the design pressure of 41.4 kPa(g) during pressure tests, and it must be durable for an exterior application and service conditions. Five candidate organic coating materials were selected for laboratory testing. As a result of the testing, a single-component elastomeric polyurethane coating was selected to be used on the dome. This paper discusses the selection process, laboratory tests and results, and the application of the polyurethane coating system to the exterior concrete dome surface. However, the main emphasis of the paper is on the laboratory evaluation of the five candidate materials. (author). 2 refs., 3 tabs., 1 fig

Edible foam based on Pickering effect of probiotic bacteria and milk proteins

DEFF Research Database (Denmark)

Yücel, Cigdem; Geng, Xiaolu; Cárdenas, Marité

2017-01-01

We report the preparation and characterization of aqueous Pickering foams using bio-particles constituted by lactic acid bacteria surface modified by oppositely charged milk proteins. Cell surface modification was shown by zeta potential measurements. Foams stabilized by bacterial Pickering bio-p...

Superhydrophobic cellulose-based bionanocomposite films from Pickering emulsions

Science.gov (United States)

Bayer, Ilker S.; Steele, Adam; Martorana, Philip J.; Loth, Eric; Miller, Lance

2009-04-01

Inherently superhydrophobic and flexible cellulose-based bionanocomposites were fabricated from solid stabilized (Pickering) emulsions. Emulsions were formed by dispersing cyclosiloxanes in water stabilized by layered silicate particles and were subsequently modified by blending into a zinc oxide nanofluid. The polymer matrix was a blend of cellulose nitrate and fluoroacrylic polymer (Zonyl 8740) precompatibilized in solution. Coatings were spray cast onto aluminum substrates from polymer blends dispersed in modified Pickering emulsions. No postsurface treatment was required to induce superhydrophobicity. Effect of antiseptic additives on bionanocomposite superhydrophobicity is also discussed. Replacing cellulose nitrate with commercial liquid bandage solutions produced identical superhydrophobic coatings.

Ergonomic Evaluation of Battery Powered Portable Cotton Picker

Science.gov (United States)

Dixit, A.; Manes, G. S.; Singh, A.; Prakash, A.; Mahal, J. S.

2012-09-01

Ergonomic evaluation of battery powered portable manual cotton picker was carried out on two subjects for three cotton varieties and was compared against manual method of picking. It is a hand operated machine and has a pair of chain with small sharp edged teeth and sprockets and is operated by a light weight 12 V battery. Cotton gets entangled with the chain and is collected and guided into the collection bag. Average heart rate, oxygen consumption, workload, energy expenditure was more in case of cotton picking by manual cotton picker as compared to manual picking for both the subjects for all three cotton variety types. Oxygen consumption varied from 0.81 to 0.97 l/min, workload varied from 36.32 to 46.16 W and energy expenditure varied from 16.83 to 20.33 kJ/min for both the subject in case of machine picking for all three cotton varieties. The maximum discomfort experienced by the subjects during picking cotton by manual cotton picker was in right wrist palm, right forearm, upper and lower back, left shoulder and in lower legs and both feet.

Wet steam turbines for CANDU-Reactors

International Nuclear Information System (INIS)

Westmacott, C.H.L.

1977-01-01

The technical characteristics of 4 wet steam turbine aggregates used in the Pickering nuclear power station are reported on along with operational experience. So far, the general experience was positive. Furthermore, plans are mentioned to use this type of turbines in other CANDU reactors. (UA) [de

Preparation of Pickering emulsions through interfacial adsorption by soft cyclodextrin nanogels

Directory of Open Access Journals (Sweden)

Shintaro Kawano

2015-11-01

Full Text Available Background: Emulsions stabilized by colloidal particles are known as Pickering emulsions. To date, soft microgel particles as well as inorganic and organic particles have been utilized as Pickering emulsifiers. Although cyclodextrin (CD works as an attractive emulsion stabilizer through the formation of a CD–oil complex at the oil–water interface, a high concentration of CD is normally required. Our research focuses on an effective Pickering emulsifier based on a soft colloidal CD polymer (CD nanogel with a unique surface-active property.Results: CD nanogels were prepared by crosslinking heptakis(2,6-di-O-methyl-β-cyclodextrin with phenyl diisocyanate and subsequent immersion of the resulting polymer in water. A dynamic light scattering study shows that primary CD nanogels with 30–50 nm diameter assemble into larger CD nanogels with 120 nm diameter by an increase in the concentration of CD nanogel from 0.01 to 0.1 wt %. The CD nanogel has a surface-active property at the air–water interface, which reduces the surface tension of water. The CD nanogel works as an effective Pickering emulsion stabilizer even at a low concentration (0.1 wt %, forming stable oil-in-water emulsions through interfacial adsorption by the CD nanogels.Conclusion: Soft CD nanogel particles adsorb at the oil–water interface with an effective coverage by forming a strong interconnected network and form a stable Pickering emulsion. The adsorption property of CD nanogels on the droplet surface has great potential to become new microcapsule building blocks with porous surfaces. These microcapsules may act as stimuli-responsive nanocarriers and nanocontainers.

The Pickering mesonet 1988 data report

International Nuclear Information System (INIS)

Salmon, J.R.; Taylor, P.A.

1989-10-01

This report describes the demonstration mesoscale meteorological monitoring network (mesonet) installed in the vicinity of the Pickering Nuclear Generating Station. It also summarizes the data collected by the network during 1988 and provides some examples of situations in which mesoscale effects dominate the local wind flow

Safe "cloudification" of large images through picker APIs.

Science.gov (United States)

Bremer, Erich; Kurc, Tahsin; Gao, Yi; Saltz, Joel; Almeida, Jonas S

2016-01-01

The "Box model" allows users with no particular training in informatics, or access to specialized infrastructure, operate generic cloud computing resources through a temporary URI dereferencing mechanism known as "drop-file-picker API" ("picker API" for sort). This application programming interface (API) was popularized in the web app development community by DropBox, and is now a consumer-facing feature of all major cloud computing platforms such as Box.com, Google Drive and Amazon S3. This reports describes a prototype web service application that uses picker APIs to expose a new, "cloudified", API tailored for image analysis, without compromising the private governance of the data exposed. In order to better understand this cross-platform cloud computing landscape, we first measured the time for both transfer and traversing of large image files generated by whole slide imaging (WSI) in Digital Pathology. The verification that there is extensive interconnectivity between cloud resources let to the development of a prototype software application that exposes an image-traversing REST API to image files stored in any of the consumer-facing "boxes". In summary, an image file can be upload/synchronized into a any cloud resource with a file picker API and the prototype service described here will expose an HTTP REST API that remains within the safety of the user's own governance. The open source prototype is publicly available at sbu-bmi.github.io/imagebox. Availability The accompanying prototype application is made publicly available, fully functional, with open source, at http://sbu-bmi.github.io/imagebox://sbu-bmi.github.io/imagebox. An illustrative webcasted use of this Web App is included with the project codebase at https://github.com/SBU-BMI/imageboxs://github.com/SBU-BMI/imagebox.

Pickering unit 1 containment leakage characterization

International Nuclear Information System (INIS)

Zakaib, G.D.

1994-01-01

Results of the design pressure test carried out on Pickering Reactor Building number 1 during late 1992 showed that the leakage rate of the building was close to the safety analysis value of 2.7% contained mass per hour at the design pressure of 41.4 kPa(g) and was significantly higher than that reported after the previous test conducted in the spring of 1987. This unexpected finding initiated the longest and the most comprehensive containment leakage investigation ever undertaken by Ontario Hydro. A thorough investigation of leakage behaviour by repeated testing, inspections, leak search and analysis was launched. The extensive leak search effort included items such as: leak source detection by soap solution application, use of ultrasonic detectors, fogging and tracer gas techniques, systematic systems isolation, thermal imaging of the exterior, and quantification of leak sites by flowmeter and bagging. Using a specially designed volumetric technique, the root cause of the problem was finally confirmed as being due to 'pressure dependent laminar leakage' through the hairline cracks in the dome concrete. Structural analysis indicated that the thermal gradients and pressure loading combined to cause the cracking early in the structure's operating history and that overall structural integrity has not been compromised. Leakage rate analysis using a new fluid mechanics model augmented by the effect of thermal strains indicated that the leakage could be significantly less under certain transient temperature gradient conditions. Several options for repairing the dome were considered by a multidisciplinary team and it was finally decided to apply a specially engineered multilayer elastomeric coating to the exterior concrete surface. When the unit was re-tested in October 1993, a dramatic ten-fold improvement in leakage rate (down to 0.25%/h at design pressure) was observed. This is lower than even the commissioning results and comparable to the performance of newer units

DeepPicker: A deep learning approach for fully automated particle picking in cryo-EM.

Science.gov (United States)

Wang, Feng; Gong, Huichao; Liu, Gaochao; Li, Meijing; Yan, Chuangye; Xia, Tian; Li, Xueming; Zeng, Jianyang

2016-09-01

Particle picking is a time-consuming step in single-particle analysis and often requires significant interventions from users, which has become a bottleneck for future automated electron cryo-microscopy (cryo-EM). Here we report a deep learning framework, called DeepPicker, to address this problem and fill the current gaps toward a fully automated cryo-EM pipeline. DeepPicker employs a novel cross-molecule training strategy to capture common features of particles from previously-analyzed micrographs, and thus does not require any human intervention during particle picking. Tests on the recently-published cryo-EM data of three complexes have demonstrated that our deep learning based scheme can successfully accomplish the human-level particle picking process and identify a sufficient number of particles that are comparable to those picked manually by human experts. These results indicate that DeepPicker can provide a practically useful tool to significantly reduce the time and manual effort spent in single-particle analysis and thus greatly facilitate high-resolution cryo-EM structure determination. DeepPicker is released as an open-source program, which can be downloaded from https://github.com/nejyeah/DeepPicker-python. Copyright © 2016 Elsevier Inc. All rights reserved.

Pickering interfacial catalysis for biphasic systems: from emulsion design to green reactions.

Science.gov (United States)

Pera-Titus, Marc; Leclercq, Loïc; Clacens, Jean-Marc; De Campo, Floryan; Nardello-Rataj, Véronique

2015-02-09

Pickering emulsions are surfactant-free dispersions of two immiscible fluids that are kinetically stabilized by colloidal particles. For ecological reasons, these systems have undergone a resurgence of interest to mitigate the use of synthetic surfactants and solvents. Moreover, the use of colloidal particles as stabilizers provides emulsions with original properties compared to surfactant-stabilized emulsions, microemulsions, and micellar systems. Despite these specific advantages, the application of Pickering emulsions to catalysis has been rarely explored. This Minireview describes very recent examples of hybrid and composite amphiphilic materials for the design of interfacial catalysts in Pickering emulsions with special emphasis on their assets and challenges for industrially relevant biphasic reactions in fine chemistry, biofuel upgrading, and depollution. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Synthesis of Covalently Cross-Linked Colloidosomes from Peroxidized Pickering Emulsions

Directory of Open Access Journals (Sweden)

Nadiya Popadyuk

2016-10-01

Full Text Available A new approach to the formation of cross-linked colloidosomes was developed on the basis of Pickering emulsions that were stabilized exclusively by peroxidized colloidal particles. Free radical polymerization and a soft template technique were used to convert droplets of a Pickering emulsion into colloidosomes. The peroxidized latex particles were synthesized in the emulsion polymerization process using amphiphilic polyperoxide copolymers poly(2-tert-butylperoxy-2-methyl-5-hexen-3-ine-co-maleic acid (PM-1-MAc or poly[N-(tert-butylperoxymethylacrylamide]-co-maleic acid (PM-2-MAc, which were applied as both initiators and surfactants (inisurfs. The polymerization in the presence of the inisurfs results in latexes with a controllable amount of peroxide and carboxyl groups at the particle surface. Peroxidized polystyrene latex particles with a covalently grafted layer of inisurf PM-1-MAc or PM-2-MAc were used as Pickering stabilizers to form Pickering emulsions. A mixture of styrene and/or butyl acrylate with divinylbenzene and hexadecane was applied as a template for the synthesis of colloidosomes. Peroxidized latex particles located at the interface are involved in the radical reactions of colloidosomes formation. As a result, covalently cross-linked colloidosomes were obtained. It was demonstrated that the structure of the synthesized (using peroxidized latex particles colloidosomes depends on the amount of functional groups and pH during the synthesis. Therefore, the size and morphology of colloidosomes can be controlled by latex particle surface properties.

Slit-burst testing of cold-worked Zr-2.5 wt.% Nb pressure tubing for CANDU-PHW reactors

International Nuclear Information System (INIS)

Wilkins, B.J.S.; Barrie, J.N.; Zink, R.J.

1978-12-01

This report documents the available data on critical crack length of cold-worked Zr-2.5 wt.% Nb pressure tubing in CANDU reactors. In particular, it includes data for tubing removed from the Pickering 3 and 4 reactors. (author)

The One-Step Pickering Emulsion Polymerization Route for Synthesizing Organic-Inorganic Nanocomposite Particles

Directory of Open Access Journals (Sweden)

Kaushal Rege

2010-02-01

Full Text Available Polystyrene-silica core-shell nanocomposite particles are successfully prepared via one-step Pickering emulsion polymerization. Possible mechanisms of Pickering emulsion polymerization are addressed in the synthesis of polystyrene-silica nanocomposite particles using 2,2-azobis(2-methyl-N-(2-hydroxyethylpropionamide (VA-086 and potassium persulfate (KPS as the initiator. Motivated by potential applications of “smart” composite particles in controlled drug delivery, the one-step Pickering emulsion polymerization route is further applied to synthesize polystyrene/poly(N-isopropylacrylamide (PNIPAAm-silica core-shell nanoparticles with N-isopropylacrylamide incorporated into the core as a co-monomer. The polystyrene/PNIPAAm-silica composite nanoparticles are temperature sensitive and can be taken up by human prostate cancer (PC3-PSMA cells.

From pioneering to implementing automated blood pressure measurement in clinical practice: Thomas Pickering's legacy

DEFF Research Database (Denmark)

Stolarz-Skrzypek, Katarzyna; Thijs, Lutgarde; Wizner, Barbara

2010-01-01

Thomas G. Pickering spent most of his scientific career in carrying out research on clinical hypertension and blood pressure (BP) measurement. In our review of Pickering's seminal work, we first focused on white-coat hypertension and masked hypertension, two terms that he had introduced. Next, we...

Women cotton pickers perceptions about health hazards due to pesticide use in irrigated punjab

International Nuclear Information System (INIS)

Abbas, M.; Mehmood, I.; Bashir, A.; Hassan, S.

2015-01-01

In Pakistan, cotton crop has special importance from the perspective of largest employment generation both for males and females in the production and value chains. Cotton picking is primarily a female specific activity in all cropping zones of Pakistan. Women cotton pickers mostly belong to poor rural society involved in this labour force to feed their families. Cotton pickers in Pakistan face some serious health related problems due to heavy use of pesticides on cotton crop. The present study was designed to investigate the problem faced by women cotton pickers and their role in household decision making. Overall 150 women cotton pickers were interviewed from Bahawalnagar, Sahiwal and Vehari districts of cotton-wheat zone of the Punjab. Summary statistics of women cotton pickers' showed mean average age was 33 years and had 2.4 ears of formal schooling and 10 years of cotton picking experience. The main reasons for cotton picking reported were to reduce family financial burden (30%) followed by better access to food and resource (23%) and better education of children (21%). Majority of the respondents (97.33%) reported that the mode of payments of cotton picking was in cash and the most of the respondents (83.70%) reported that they got wages in time. Only few respondents (8.70%) were aware of health hazards due to pesticides and only 10% women wear protective clothes during cotton picking. Majority of the respondents (76%) wash their clothes after cotton picking whereas almost all the respondents wash their hand after cotton picking. The women cotton pickers faced health problem, tiredness (54.5%), mental disturbance (9.90%) and fatigue (8.00%). More than 58% women reported their involvement in household decision making regarding food and groceries while 30.6% women involved in decision about education of children. It is suggested that the female cotton pickers should be educated about the importance (in terms of disease treatment and long-run health costs

Evolution of CANDU reactor design

International Nuclear Information System (INIS)

Pon, G.A.

1978-08-01

The CANDU (CANada Deuterium Uranium) design had its begin-ings in the early 1950's with the preliminary engineering studies that led to the 20 MW(e) NPD (Nuclear Power Demonstration) and the 200 MW(e) Douglas Point station . The next decade saw the first operation of both these stations and the commitment of the 2000 MW(e) Pickering and 3000 MW(e) Bruce plants. The present decade has witnessed the excellent performance of Pickering and Bruce and commitments to construct Gentilly-2, Cordoba, Pt. Lepreau, Wolsung, Pickering B, Bruce B and Darlington. In most cases, successive CANDU designs have meant an increase in plant output. Evolutionary developments have been made to fit the requirements of higher ratings and sizes, new regulations, better reliability and maintainability and lower costs. These changes, which are described system by system, have been introduced in the course of engineering parallel reactor projects with overlapping construction schedules -circumstances which ensure close contact with the practical realities of economics, manufacturing functions, construction activities and performance in commissioning. Features for one project furnished alternative concepts for others still on the drawing board and the experience gained in the first application yielded a sound basis for its re-use in succeeding projects. Thus the experiences gained in NPD, Douglas Point, Gentilly-1 and KANUPP have contributed to Pickering and Bruce, which in turn have contributed to the design of Gentilly-2. (author)

Safe “cloudification” of large images through picker APIs

Science.gov (United States)

Bremer, Erich; Kurc, Tahsin; Gao, Yi; Saltz, Joel; Almeida, Jonas S

2016-01-01

The “Box model” allows users with no particular training in informatics, or access to specialized infrastructure, operate generic cloud computing resources through a temporary URI dereferencing mechanism known as “drop-file-picker API” (“picker API” for sort). This application programming interface (API) was popularized in the web app development community by DropBox, and is now a consumer-facing feature of all major cloud computing platforms such as Box.com, Google Drive and Amazon S3. This reports describes a prototype web service application that uses picker APIs to expose a new, “cloudified”, API tailored for image analysis, without compromising the private governance of the data exposed. In order to better understand this cross-platform cloud computing landscape, we first measured the time for both transfer and traversing of large image files generated by whole slide imaging (WSI) in Digital Pathology. The verification that there is extensive interconnectivity between cloud resources let to the development of a prototype software application that exposes an image-traversing REST API to image files stored in any of the consumer-facing “boxes”. In summary, an image file can be upload/synchronized into a any cloud resource with a file picker API and the prototype service described here will expose an HTTP REST API that remains within the safety of the user’s own governance. The open source prototype is publicly available at sbu-bmi.github.io/imagebox. Availability The accompanying prototype application is made publicly available, fully functional, with open source, at http://sbu-bmi.github.io/imagebox://sbu-bmi.github.io/imagebox. An illustrative webcasted use of this Web App is included with the project codebase at https://github.com/SBU-BMI/imageboxs://github.com/SBU-BMI/imagebox. PMID:28269829

Pickering Unit 1 chemical cleaning

International Nuclear Information System (INIS)

Smee, J.L.; Fiola, R.J.; Brennenstuhl, K.R.; Zerkee, D.D.; Daniel, C.M.

1995-01-01

The secondary sides of all 12 boilers at Pickering Unit 1 were chemically cleaned in 1994 by the team of Ontario Hydro, B and W International (Cambridge, Ontario) and B and W Nuclear Technologies (Lynchburg, Virginia). A multi-step EPRI/SGOG process was employed in a similar manner to previous clearings at Units 5 and 6 in 1992 and 1993, respectively. A major innovation with the Unit 1 cleaning was the incorporation of a crevice cleaning step, the first time this had been done on Ontario Hydro plants. In addition, six boilers were cleaned in parallel compared to three at a time in previous Pickering cleanings. This significantly reduced cleaning time. A total of 6,770 kg of sludge was removed through direct chemical dissolution. It consisted of 66% iron/nickel oxides and 28% copper metal. A total of 1,600,000 L (420,000 US gallons) of liquid waste was produced. It was processed through the spent solvent treatment facility located at the Bruce Nuclear Power Development site. Visual inspection performed after the cleaning indicated that the crevices between the boiler tubes and the tube support structure were completely clear of deposit and the general condition of the tubing and lattice bars appeared to be in 'as new' condition. (author)

AECB staff annual report of Pickering NGS for the year 1991

International Nuclear Information System (INIS)

1992-11-01

The AECB Pickering project staff, in cooperation with AECB staff in Ottawa, monitor the operation of Pickering NGS-A units 1-4 and Pickering NGS-B units 5-8 to ensure that Ontario Hydro operates the station in compliance with the licensing and safety requirements of the Atomic Energy Control Board. This report presents the review of licensing issues and station performance during 1991. Improvement over 1990 station operation occurred in the following areas: availability of special safety systems; reduction of the station external dose; reorganization of station management to improve focus; station chemistry; housekeeping and material condition; fuel handling capability; training of operators and maintenance staff. However, little change occurred and improvement is still needed in the following: compliance with operating licence; system surveillance program; station maintenance; environmental qualification; radiation emergency response; fire and rescue emergency response; limited capability to predict and prevent equipment failures such as the boiler tube failure on unit 5. (L.L.)

Warehouse order-picking process. Order-picker routing problem

Directory of Open Access Journals (Sweden)

E. V. Korobkov

2015-01-01

Full Text Available This article continues “Warehouse order-picking process” cycle and describes order-picker routing sub-problem of a warehouse order-picking process. It draws analogies between the orderpickers’ routing problem and traveling salesman’s problem, shows differences between the standard problem statement of a traveling salesman and routing problem of warehouse orderpickers, and gives the particular Steiner’s problem statement of a traveling salesman.Warehouse layout with a typical order is represented by a graph, with some its vertices corresponding to mandatory order-picker’s visits and some other ones being noncompulsory. The paper describes an optimal Ratliff-Rosenthal algorithm to solve order-picker’s routing problem for the single-block warehouses, i.e. warehouses with only two crossing aisles, defines seven equivalent classes of partial routing sub-graphs and five transitions used to have an optimal routing sub-graph of a order-picker. An extension of optimal Ratliff-Rosenthal order-picker routing algorithm for multi-block warehouses is presented and also reasons for using the routing heuristics instead of exact optimal algorithms are given. The paper offers algorithmic description of the following seven routing heuristics: S-shaped, return, midpoint, largest gap, aisle-by-aisle, composite, and combined as well as modification of combined heuristics. The comparison of orderpicker routing heuristics for one- and two-block warehouses is to be described in the next article of the “Warehouse order-picking process” cycle.

Development and characterization of novel antimicrobial bilayer films based on Polylactic acid (PLA)/Pickering emulsions.

Science.gov (United States)

Zhu, Jun-You; Tang, Chuan-He; Yin, Shou-Wei; Yang, Xiao-Quan

2018-02-01

Biodegradable food packaging is sustainable and has a great application prospect. PLA is a promising alternative for petroleum-derived polymers. However, PLA packaging suffers from poor barrier properties compared with petroleum-derived ones. To address this issue, we designed bilayer films based on PLA and Pickering emulsions. The formed bilayer films were compact and uniform and double layers were combined firmly. This strategy enhanced mechanical resistance, ductility and moisture barrier of Pickering emulsion films, and concomitantly enhanced the oxygen barrier for PLA films. Thymol loadings in Pickering emulsion layer endowed them with antimicrobial and antioxidant activity. The release profile of thymol was well fitted with Fick's second law. The antimicrobial activity of the films depended on film types, and Pickering emulsion layer presented larger inhibition zone than PLA layer, hinting that the films possessed directional releasing role. This study opens a promising route to fabricate bilayer architecture creating synergism of each layer. Copyright © 2017 Elsevier Ltd. All rights reserved.

Review of Ontario Hydro Pickering 'A' and Bruce 'A' nuclear generating stations' accident analyses

International Nuclear Information System (INIS)

Serdula, K.J.

1988-01-01

Deterministic safety analysis for the Pickering 'A' and Bruce 'A' nuclear generating stations were reviewed. The methodology used in the evaluation and assessment was based on the concept of 'N' critical parameters defining an N-dimensional safety parameter space. The reviewed accident analyses were evaluated and assessed based on their demonstrated safety coverage for credible values and trajectories of the critical parameters within this N-dimensional safety parameter space. The reported assessment did not consider probability of occurrence of event. The reviewed analyses were extensive for potential occurrence of accidents under normal steady-state operating conditions. These analyses demonstrated an adequate assurance of safety for the analyzed conditions. However, even for these reactor conditions, items have been identified for consideration of review and/or further study, which would provide a greater assurance of safety in the event of an accident. Accident analyses based on a plant in a normal transient operating state or in an off-normal condition but within the allowable operating envelope are not as extensive. Improvements in demonstrations and/or justifications of safety upon potential occurrence of accidents would provide further assurance of adequacy of safety under these conditions. Some events under these conditions have not been analyzed because of their judged low probability; however, accident analyses in this area should be considered. Recommendations are presented relating to these items; it is also recommended that further study is needed of the Pickering 'A' special safety systems

Pickering emulsions for skin decontamination.

Science.gov (United States)

Salerno, Alicia; Bolzinger, Marie-Alexandrine; Rolland, Pauline; Chevalier, Yves; Josse, Denis; Briançon, Stéphanie

2016-08-01

This study aimed at developing innovative systems for skin decontamination. Pickering emulsions, i.e. solid-stabilized emulsions, containing silica (S-PE) or Fuller's earth (FE-PE) were formulated. Their efficiency for skin decontamination was evaluated, in vitro, 45min after an exposure to VX, one of the most highly toxic chemical warfare agents. Pickering emulsions were compared to FE (FE-W) and silica (S-W) aqueous suspensions. PE containing an oil with a similar hydrophobicity to VX should promote its extraction. All the formulations reduced significantly the amount of VX quantified on and into the skin compared to the control. Wiping the skin surface with a pad already allowed removing more than half of VX. FE-W was the less efficient (85% of VX removed). The other formulations (FE-PE, S-PE and S-W) resulted in more than 90% of the quantity of VX removed. The charge of particles was the most influential factor. The low pH of formulations containing silica favored electrostatic interactions of VX with particles explaining the better elimination from the skin surface. Formulations containing FE had basic pH, and weak interactions with VX did not improve the skin decontamination. However, these low interactions between VX and FE promote the transfer of VX into the oil droplets in the FE-PE. Copyright © 2016 Elsevier B.V. All rights reserved.

The number of pickers and stock-keeping unit arrangement on a uni-directional picking line

Directory of Open Access Journals (Sweden)

Hagspihl, Robert

2014-10-01

Full Text Available The order picking process is often the single largest expense in a distribution centre (DC. The DC considered in this paper uses a picking line configuration to perform order picking. The number of pickers in a picking line, and the initial arrangement of stock-keeping units (SKUs, are two important factors that affect the total completion time of the picking lines. In this paper, the picking line configuration is simulated with an agent-based approach to describe the behaviour of an individual picker. The simulation is then used to analyse the effect of the number of pickers and the SKU arrangement. Verification and validation of this model shows that the model represents the real-world picking line to a satisfactory degree. Marginal analysis (MA was chosen to determine a ‘good’ number of pickers by means of the simulation model. A look-up table is presented to provide decision support for the choice of a ‘good’ number of pickers to improve completion times of the picking line, for the properties of a specific picking line. The initial SKU arrangement on a picking line is shown to be a factor that can affect the level of picker congestion and the total completion time. The greedy ranking and partitioning (GRP and organ pipe arrangement (OPA techniques from the literature, as well as the historical SKU arrangements used by the retailer under consideration, were compared with the proposed classroom discipline heuristic (CDH for SKU arrangement. It was found that the CDH provides an more even spread of SKUs that are picked most frequently, thus decreasing congestion and total completion time.

Exploiting the pliability and lateral mobility of Pickering emulsion for enhanced vaccination

Science.gov (United States)

Xia, Yufei; Wu, Jie; Wei, Wei; Du, Yiqun; Wan, Tao; Ma, Xiaowei; An, Wenqi; Guo, Aiying; Miao, Chunyu; Yue, Hua; Li, Shuoguo; Cao, Xuetao; Su, Zhiguo; Ma, Guanghui

2018-02-01

A major challenge in vaccine formulations is the stimulation of both the humoral and cellular immune response for well-defined antigens with high efficacy and safety. Adjuvant research has focused on developing particulate carriers to model the sizes, shapes and compositions of microbes or diseased cells, but not antigen fluidity and pliability. Here, we develop Pickering emulsions--that is, particle-stabilized emulsions that retain the force-dependent deformability and lateral mobility of presented antigens while displaying high biosafety and antigen-loading capabilities. Compared with solid particles and conventional surfactant-stabilized emulsions, the optimized Pickering emulsions enhance the recruitment, antigen uptake and activation of antigen-presenting cells, potently stimulating both humoral and cellular adaptive responses, and thus increasing the survival of mice upon lethal challenge. The pliability and lateral mobility of antigen-loaded Pickering emulsions may provide a facile, effective, safe and broadly applicable strategy to enhance adaptive immunity against infections and diseases.

Fuel deposits, chemistry and CANDU® reactor operation

International Nuclear Information System (INIS)

Roberts, J.G.

2014-01-01

'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU® reactor, the first being the Nuclear Power Demonstration - 2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channelled to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5. The difference being that during 'hot conditioning' of CANDU® heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

Heterogeneous Pd catalysts as emulsifiers in Pickering emulsions for integrated multistep synthesis in flow chemistry.

Science.gov (United States)

Hiebler, Katharina; Lichtenegger, Georg J; Maier, Manuel C; Park, Eun Sung; Gonzales-Groom, Renie; Binks, Bernard P; Gruber-Woelfler, Heidrun

2018-01-01

Within the "compartmentalised smart factory" approach of the ONE-FLOW project the implementation of different catalysts in "compartments" provided by Pickering emulsions and their application in continuous flow is targeted. We present here the development of heterogeneous Pd catalysts that are ready to be used in combination with biocatalysts for catalytic cascade synthesis of active pharmaceutical ingredients (APIs). In particular, we focus on the application of the catalytic systems for Suzuki-Miyaura cross-coupling reactions, which is the key step in the synthesis of the targeted APIs valsartan and sacubitril. An immobilised enzyme will accomplish the final product formation via hydrolysis. In order to create a large interfacial area for the catalytic reactions and to keep the reagents separated until required, the catalyst particles are used to stabilise Pickering emulsions of oil and water. A set of Ce-Sn-Pd oxides with the molecular formula Ce 0.99- x Sn x Pd 0.01 O 2-δ ( x = 0-0.99) has been prepared utilising a simple single-step solution combustion method. The high applicability of the catalysts for different functional groups and their minimal leaching behaviour is demonstrated with various Suzuki-Miyaura cross-coupling reactions in batch as well as in continuous flow employing the so-called "plug & play reactor". Finally, we demonstrate the use of these particles as the sole emulsifier of oil-water emulsions for a range of oils.

Novel carboxymethyl cellulose-polyvinyl alcohol blend films stabilized by Pickering emulsion incorporation method.

Science.gov (United States)

Fasihi, Hadi; Fazilati, Mohammad; Hashemi, Mahdi; Noshirvani, Nooshin

2017-07-01

The aim of this study was to investigate the possibility of increasing the antimicrobial and antioxidant properties of biodegradable active films stabilized via Pickering emulsions. The blend films were prepared from carboxymethyl cellulose (CMC) and polyvinyl alcohol (PVA), emulsified with oleic acid (OL) and incorporated with rosemary essential oil (REO). Formation of Pickering emulsion was confirmed by scanning electron microscopy (SEM), optical microscopy, mean droplet size and emulsion stability. Morphological, optical, physical, mechanical, thermal, antifungal and antioxidant properties of the films incorporated with different concentrations of REO (0.5, 1.5 and 3%) were determined. The results showed an increase in UV absorbance and elongation at break but, a decrease in tensile strength and thermal stability of the films. Interestingly, films containing REO exhibited considerable antioxidant and antimicrobial properties. In vitro microbial tests exhibited 100% fungal inhibition against Penicillium digitatum in the films containing 3% REO. In addition, no fungal growth were observed after 60days of storage at 25°C in bread slices were stored with active films incorporated with 3% REO, could attributed to the slow and regular release of REO caused by Pickering emulsions. The results of this study suggest that Pickering emulsion is a very promising method, which significantly affects antioxidant and antimicrobial activities of the films. Copyright © 2017 Elsevier Ltd. All rights reserved.

Moderator inlet line hanger replacement for Pickering nuclear power station

International Nuclear Information System (INIS)

Kirkpatrick, R.A.; Bowman, J.M.; Symmons, W.R.; El-Nesr, S.

1988-01-01

Ontario Hydro's Pickering Nuclear Generating Station (PNGS), Units 1 and 2 were shutdown for large scale fuel channel replacement. Other nonroutine inspection and maintenance activities were performed to determine the overall condition of the units and it was seen that a moderator inlet line hanger (identified as HR-29) had failed in both units. Subsequent inspections during planned maintenance outages of Pickering NGS Units 3 and 4 revealed that hanger HR-29 had failed and required replacement. A research program was conducted to find a suitable technique. These problems included accessing tooling through small inspection ports, manipulating tooling from a significant distance and the high radiation fields within the vault. This paper describes the program undertaken to replace hanger HR-29. (author)

Preference of multi-walled carbon nanotube (MWCNT) to single-walled carbon nanotube (SWCNT) and activated carbon for preparing silica nanohybrid pickering emulsion for chemical enhanced oil recovery (C-EOR)

Energy Technology Data Exchange (ETDEWEB)

AfzaliTabar, M. [Department of Chemistry, Islamic Azad University Branch of Tehran North, Tehran (Iran, Islamic Republic of); Alaei, M., E-mail: alaiem@ripi.ir [Nanotechnology Research Center, Research Institute of Petroleum Industry (RIPI), Tehran (Iran, Islamic Republic of); Ranjineh Khojasteh, R.; Motiee, F. [Department of Chemistry, Islamic Azad University Branch of Tehran North, Tehran (Iran, Islamic Republic of); Rashidi, A.M. [Nanotechnology Research Center, Research Institute of Petroleum Industry (RIPI), Tehran (Iran, Islamic Republic of)

2017-01-15

The aim of this research was to determine the best nano hybrid that can be used as a Pickering emulsion Chemical Enhanced Oil Recovery (C-EOR). Therefore, we have prepared different carbon structures nano hybrids with SiO{sub 2} nano particles with different weight percent using sol-gel method. The as-prepared nano materials were characterized with X-Ray Diffraction (XRD), Field Emission Scanning Electron Microscopy (FE-SEM) and Thermal Gravimetric Analysis (TGA). Pickering emulsions of these nanohybrids were prepared at pH=7 in ambient temperature and with distilled water. Stability of the mentioned Pickering emulsions was controlled for one month. Emulsion phase morphology was investigated using optical microscopic imaging. Evaluation results demonstrated that the best sample is the 70% MWCNT/SiO{sub 2} nanohybrid. Stability of the selected nanohybrid (70% MWCNT/SiO{sub 2} nanohybrid) was investigated by alteration of salinity, pH and temperature. Results showed that the mentioned Pickering emulsion has very good stability at 0.1%, 1% salinity, moderate and high temperature (25 °C and 90 °C) and neutral and alkaline pH (7, 10) that is suitable for the oil reservoirs conditions. The effect of the related nano fluid on the wettability of carbonate rock was investigated by measuring the contact angle and interfacial tension. Results show that the nanofluid could significantly change the wettability of the carbonate rock from oil wet to water wet and can decrease the interfacial tension. Therefore, the 70% MWCNT/SiO{sub 2} nanohybrid Pickering emulsion can be used for Chemical Enhanced Oil Recovery (C-EOR).

Preference of multi-walled carbon nanotube (MWCNT) to single-walled carbon nanotube (SWCNT) and activated carbon for preparing silica nanohybrid pickering emulsion for chemical enhanced oil recovery (C-EOR)

International Nuclear Information System (INIS)

AfzaliTabar, M.; Alaei, M.; Ranjineh Khojasteh, R.; Motiee, F.; Rashidi, A.M.

2017-01-01

The aim of this research was to determine the best nano hybrid that can be used as a Pickering emulsion Chemical Enhanced Oil Recovery (C-EOR). Therefore, we have prepared different carbon structures nano hybrids with SiO 2 nano particles with different weight percent using sol-gel method. The as-prepared nano materials were characterized with X-Ray Diffraction (XRD), Field Emission Scanning Electron Microscopy (FE-SEM) and Thermal Gravimetric Analysis (TGA). Pickering emulsions of these nanohybrids were prepared at pH=7 in ambient temperature and with distilled water. Stability of the mentioned Pickering emulsions was controlled for one month. Emulsion phase morphology was investigated using optical microscopic imaging. Evaluation results demonstrated that the best sample is the 70% MWCNT/SiO 2 nanohybrid. Stability of the selected nanohybrid (70% MWCNT/SiO 2 nanohybrid) was investigated by alteration of salinity, pH and temperature. Results showed that the mentioned Pickering emulsion has very good stability at 0.1%, 1% salinity, moderate and high temperature (25 °C and 90 °C) and neutral and alkaline pH (7, 10) that is suitable for the oil reservoirs conditions. The effect of the related nano fluid on the wettability of carbonate rock was investigated by measuring the contact angle and interfacial tension. Results show that the nanofluid could significantly change the wettability of the carbonate rock from oil wet to water wet and can decrease the interfacial tension. Therefore, the 70% MWCNT/SiO 2 nanohybrid Pickering emulsion can be used for Chemical Enhanced Oil Recovery (C-EOR).

Pickering Emulsions for Food Applications: Background, Trends, and Challenges

NARCIS (Netherlands)

Berton-Carabin, C.C.; Schroën, C.G.P.H.

2015-01-01

Particle-stabilized emulsions, also referred to as Pickering emulsions, have garnered exponentially increasing interest in recent years. This has also led to the first food applications, although the number of related publications is still rather low. The involved stabilization mechanisms are

9 CFR 78.7 - Brucellosis reactor cattle.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Brucellosis reactor cattle. 78.7... Restrictions on Interstate Movement of Cattle Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a) Destination. Brucellosis reactor cattle may be moved interstate only for immediate slaughter as follows: (1...

Tailoring the Wettability of Colloidal Particles for Pickering Emulsions via Surface Modification and Roughness

Directory of Open Access Journals (Sweden)

Meina Xiao

2018-06-01

Full Text Available Pickering emulsions are water or oil droplets that are stabilized by colloidal particles and have been intensely studied since the late 90s. The surfactant-free nature of these emulsions has little adverse effects such as irritancy and contamination of environment and typically exhibit enhanced stability compared to surfactant-stabilized emulsions. Therefore, they offer promising applications in cosmetics, food science, controlled release, and the manufacturing of microcapsules and porous materials. The wettability of the colloidal particles is the main parameter determining the formation and stability of Pickering emulsions. Tailoring the wettability by surface chemistry or surface roughness offers considerable scope for the design of a variety of hybrid nanoparticles that may serve as novel efficient Pickering emulsion stabilizers. In this review, we will discuss the recent advances in the development of surface modification of nanoparticles.

Particle Shape Anisotropy in Pickering Emulsions: Cubes and Peanuts

NARCIS (Netherlands)

de Folter, J.W.J.; Hutter, E.M.; Castillo, S.I.R.; Klop, K.E.; Philipse, A.P.; Kegel, W.K.

2014-01-01

We have investigated the effect of particle shape in Pickering emulsions by employing, for the first time, cubic and peanut-shaped particles. The interfacial packing and orientation of anisotropic microparticles are revealed at the single-particle level by direct microscopy observations. The uniform

Seismic assessment of the Pickering pressure relief duct

International Nuclear Information System (INIS)

Ghobarah, A.

1995-05-01

The objectives of the study are to examine the structural response of the Pickering pressure relief duct when subjected to earthquake ground motion and to estimate the seismic withstand capacity of various components of the structural system on the basis of performance criteria consistent with the safety function of the duct. (author). 24 refs., 16 tabs., 31 figs

Aquivion Perfluorosulfonic Superacid as an Efficient Pickering Interfacial Catalyst for the Hydrolysis of Triglycerides.

Science.gov (United States)

Shi, Hui; Fan, Zhaoyu; Hong, Bing; Pera-Titus, Marc

2017-09-11

Rational design of the surface properties of heterogeneous catalysts can boost the interfacial activity in biphasic reactions through the generation of Pickering emulsions. This concept, termed Pickering interfacial catalysis (PIC), has shown promising credentials in acid-catalyzed transesterification, ester hydrolysis, acetalization, etherification, and alkylation reactions. PIC has now been applied to the efficient, solvent-free hydrolysis of the triglyceride glyceryl trilaurate to lauric acid, catalyzed by Aquivion perfluorosulfonic superacid at mild conditions (100 °C and ambient pressure). © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Study of the time effect on the strength of cell-cell adhesion force by a novel nano-picker

Energy Technology Data Exchange (ETDEWEB)

Shen, Yajing, E-mail: shen@robo.mein.naogya-u.ac.jp [Dept. of Micro-Nano Systems Engineering, Nagoya University, Nagoya 464-8603 (Japan); Nakajima, Masahiro [Center for Micro-Nano Mechatronics, Nagoya University, Nagoya 464-8603 (Japan); Kojima, Seiji; Homma, Michio [Division of Biological Science, Nagoya University, Nagoya 464-8603 (Japan); Fukuda, Toshio [Dept. of Micro-Nano Systems Engineering, Nagoya University, Nagoya 464-8603 (Japan); Center for Micro-Nano Mechatronics, Nagoya University, Nagoya 464-8603 (Japan)

2011-06-03

Highlights: {yields} A nano-picker is developed for single cell adhesion force measurement. {yields} The adhesion of picker-cell has no influence to the cell-cell measurement result. {yields} Cell-cell adhesion force has a rise at the first few minutes and then becomes constant. -- Abstract: Cell's adhesion is important to cell's interaction and activates. In this paper, a novel method for cell-cell adhesion force measurement was proposed by using a nano-picker. The effect of the contact time on the cell-cell adhesion force was studied. The nano-picker was fabricated from an atomic force microscopy (AFM) cantilever by nano fabrication technique. The cell-cell adhesion force was measured based on the deflection of the nano-picker beam. The result suggests that the adhesion force between cells increased with the increasing of contact time at the first few minutes. After that, the force became constant. This measurement methodology was based on the nanorobotic manipulation system inside an environmental scanning electron microscope. It can realize both the observation and manipulation of a single cell at nanoscale. The quantitative and precise cell-cell adhesion force result can be obtained by this method. It would help us to understand the single cell interaction with time and would benefit the research in medical and biological fields potentially.

Measurement and analysis of pressure tube elongation in the Douglas Point reactor

International Nuclear Information System (INIS)

Causey, A.R.; MacEwan, S.R.; Jamieson, H.C.; Mitchell, A.B.

1980-02-01

Elongations of zirconium alloy pressure tubes in CANDU reactors, which occur as a result of neutron-irradiation-induced creep and growth, have been measured over the past 6 years, and the consequences of thses elongations have recently been analysed. Elongation rates, previously deduced from extensive measurements of elongations of cold-worked Zircaloy-2 pressure tubes in the Pickering reactors, have been modified to apply to the pressure tubes in the Douglas Point (DP) reactor by taking into account measured diffences in texture and dislocation density. Using these elongation rates, and structural data unique to the DP reactor, the analysis predicts elongation behaviour which is in good agreement with pressure tube elongations measured during the ten years of reactor operation. (Auth)

Condition based maintenance pilot projects at Pickering ND

International Nuclear Information System (INIS)

Zemdegs, R.T.

1995-01-01

Ontario Hydro has recognized that the approaches to maintenance have undergone significant changes to the past decades. The traditional break down maintenance approach has been replaced by preventative maintenance and more recently, by condition based maintenance. The nuclear plants of Ontario Hydro have evaluated on a number of alternative programs to improve their maintenance effectiveness and to reduce costs, including Reliability Centred Maintenance (RCM), call-up review, component-based PM programs, analysis of failure history and so on. Pickering ND (nuclear division) and Ontario Hydro's Nuclear Technologies Services Division, have embarked on a Condition Based Maintenance pilot project to address the above issues as a breakthrough solution for smarter maintenance. The Condition Based Maintenance pilot project will demonstrate an end-to-end process utilizing a Reliability Centred Maintenance structured approach to re-engineer and redefine the existing maintenance programs. The project emphasizes on-condition maintenance where justified, and utilizes an information management tool to provide the required records keeping and analysis infrastructure. This paper briefly describes the planned maintenance model at Pickering ND used to guide the CBM pilot, and an overview of the methodology used to develop on-condition equipment indicators as part of a re-engineered maintenance plan

Picker versus stripper harvesters on the High Plains of Texas

Science.gov (United States)

A break even analysis based on NPV was conducted to compare picker-based and stripper-based harvest systems with and without field cleaners. Under no conditions analyzed was the NPV of a stripper system without a field cleaner greater than a stripper system with a field cleaner. Break even curves re...

Pulse picker for synchrotron radiation driven by a surface acoustic wave.

Science.gov (United States)

Vadilonga, Simone; Zizak, Ivo; Roshchupkin, Dmitry; Petsiuk, Andrei; Dolbnya, Igor; Sawhney, Kawal; Erko, Alexei

2017-05-15

A functional test for a pulse picker for synchrotron radiation was performed at Diamond Light Source. The purpose of a pulse picker is to select which pulse from the synchrotron hybrid-mode bunch pattern reaches the experiment. In the present work, the Bragg reflection on a Si/B4C multilayer was modified using surface acoustic wave (SAW) trains. Diffraction on the SAW alters the direction of the x rays and it can be used to modulate the intensity of the x rays that reach the experimental chamber. Using electronic modulation of the SAW amplitude, it is possible to obtain different scattering conditions for different x-ray pulses. To isolate the single bunch, the state of the SAW must be changed in the short time gap between the pulses. To achieve the necessary time resolution, the measurements have been performed in conical diffraction geometry. The achieved time resolution was 120 ns.

AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1996

International Nuclear Information System (INIS)

1997-06-01

The Atomic Energy Control Board is the independent federal agency that controls all nuclear activities in Canada. A major use of nuclear energy in Canada is electricity production. The AECB assesses every station's performance against legal requirements, including the conditions in the operating licence. Each station is inspected and all aspects of the station's operation and management is reviewed. This report is the AECB staff assessment of reactor safety at the Pickering A and B Generating Stations for 1996. PNGS-A and PNGS-B operated safely during 1996. Although the risk to the workers and the public is low, major safety related changes are necessary at the stations and the sustainability of those changes needs to be demonstrated. Improvement is needed by Ontario Hydro in meeting the time limits for reporting reportable events. Ontario Hydro's follow up to events and causal factor analyses continue to need improvements. Improvements are needed to operational safety and reactor maintenance at both A and B. There are signs of improvement through Ontario Hydro's plan for recovery, and in station management changes. There also appears to be commitment to safety expressed at the highest level of the utility

Fuel deposits, chemistry and CANDU reactor operation

International Nuclear Information System (INIS)

Roberts, J.G.

2013-01-01

'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU reactor, the first being the Nuclear Power Demonstration-2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channel led to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5, and subsequently utilized for each CANDU unit since. The difference being that during 'hot conditioning' of CANDU heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

Cobalt-60 control in Ontario Hydro reactors

International Nuclear Information System (INIS)

Lacy, C.S.

1988-01-01

This paper discusses the impact of specifying reduced Cobalt-59 in the primary heat transport circuit materials of construction on the radiation fields developed around the primary circuit. An eight-fold reduction in steam generator radiation fields due to Cobalt-60 has been observed for two identical sets of reactors, one with and one without Cobalt-59 control. The comparison is between eight reactors at the Pickering Nuclear Generating Station (PNGS). Units 5 to 8 (PNGS-B) are identical to Units 1 to 4 (PNGS-A) except that PNGS-B has reduced impurity Cobalt-59 in the alloys of construction and a reduced use of stellite. The effects of chemistry control are also discussed

Pipe support program at Pickering

International Nuclear Information System (INIS)

Sahazizian, L.A.; Jazic, Z.

1997-01-01

This paper describes the pipe support program at Pickering. The program addresses the highest priority in operating nuclear generating stations, safety. We present the need: safety, the process: managed and strategic, and the result: assurance of critical piping integrity. In the past, surveillance programs periodically inspected some systems, equipment, and individual components. This comprehensive program is based on a managed process that assesses risk to identify critical piping systems and supports and to develop a strategy for surveillance and maintenance. The strategy addresses all critical piping supports. Successful implementation of the program has provided assurance of critical piping and support integrity and has contributed to decreasing probability of pipe failure, reducing risk to worker and public safety, improving configuration management, and reducing probability of production losses. (author)

Cobalt-60 production in CANDU power reactors

International Nuclear Information System (INIS)

Slack, J.; Norton, J.L.; Malkoske, G.R.

2003-01-01

therapy machines. Today the majority of the cancer therapy cobalt-60 sources used in the world are manufactured using material from the NRU reactor in Chalk River. The same technology that was used for producing cobalt-60 in a research reactor was then adapted and transferred for use in a CANDU power reactor. In the early 1970s, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production was initiated in the four Pickering A CANDU reactors located east of Toronto. This was the first full scale production of millions of curies of cobalt-60 per year. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology in additional CANDUs. Over the years MDS Nordion has partnered with CANDU reactor owners to produce cobalt-60 at various sites. CANDU reactors that have, or are still producing cobalt-60, include Pickering A, Pickering B, Gentilly 2, Embalse in Argentina, and Bruce B. In conclusion, the technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and Atomic Energy of Canada, has been safely, economically and successfully employed in CANDU reactors with over 195 reactor years of production. Today over forty percent of the world's disposable medical supplies are made safer through sterilization using cobalt-60 sources from MDS Nordion. Over the past 40 years, MDS Nordion with its CANDU reactor owner partners, has safely and reliably shipped more than 500 million curies of cobalt-60 sources to customers around the world. MDS Nordion is presently adding three more CANDU power reactors to its supply chain. These three additional cobalt producing CANDU's will help supplement the ability of the health care industry to provide safe, sterile, medical disposable products to people around the world. As new applications for cobalt-60 are identified, and the demand for bulk cobalt-60 increases, MDS Nordion and AECL

Automatic pickers performances in the case of the Emilia sequence of May-June 2012.

Science.gov (United States)

Tiberi, Lara; Spallarossa, Daniele; Costa, Giovanni

2013-04-01

The automatic processing of seismic data, whether for real-time seismic warning system or to reprocessing large amount of seismic recordings, is increasingly being demanded by seismologists especially in case of emergency as for the Emilia sequence in may-june 2012. In this study is presented a comparison between the AutoPicker (DipTeRiS, University of Genova) a new method used for automatic accurate onset phase picking for both P and S wave arrival based on the Akaike's information criterion (AIC), a solid and tested picker as the STA/LTA in Antelope software and the manual pickings. In order to construct the database used for the relocation of Emilia sequence, the RAN strong motion database has been merged with the available velocity and acceleration data extracted from the EIDA database (European Integrated Data Archive) and velocity data recorded by the Southeastern Alps Integrated Network (DMG, OGS, ARSO and ZAMG). The fault system of the Emilia earthquake area is complex and it is not easy to assess which fault has moved. A precise localization of the sequence is essential. The manual pickings, the equivalent locations and the choice of the most appropriate velocity model ("Iside") used in this study are the results of a work done in collaboration with Università di Chieti and DPC, not described here. The main problem of the AutoPicker and Antelope software is to discriminate events that occur very close to each other in time. The best way to solve that issue is choosing the best setup of both techniques to minimize the problem. Then we would like to implement the AutoPicker technique developed by Prof. Spallarossa on the Antelope system routinely used by UTS-DMG for the real-time data analysis.

A Study of Micro Finance: Special Reference to Female Waste Pickers in Pimpri Chinchwad Area in Pune

OpenAIRE

Hebalkar, Dr. Rashmi; Sharma, Meena Sunildutt

2013-01-01

Female waste pickers are the neglected section of urban women who are struggling to make ends meet, in an occupation which is hazardous for health, and are contributing to the welfare of society, without realizing it, through collecting waste and sending it forward for recycling. These women may be poorly educated but at least some of them have been unionized and their union attempts to improve their condition. Despite the existence of KKPKP union, there are female waste pickers who have not ...

Tritium releases from the Pickering Nuclear Generating Station and birth defects and infant mortality in nearby communities 1971-1988

International Nuclear Information System (INIS)

Johnson, K.C.; Rouleau, J.

1991-10-01

This study was commissioned to examine whether there were elevated rates of stillbirth, birth defects, or death in the first year of life between 1971 and 1988 among offspring of residents of communities within a 25-kilometre radius of the Pickering Nuclear Generating Station. The study was also to investigate whether there were any statistical associations between the monthly airborne or waterborne tritium emissions from the Pickering Nuclear Generating Station and the rates of these reproductive outcomes. Overall analysis did not support a hypothesis of increased rates of stillbirths, neonatal mortality or infant mortality near the Pickering Nuclear Generating Station, or a hypothesis of increased birth prevalence of birth defects for 21 of 22 diagnostic categories. The prevalence of Down Syndrome was elevated in both Pickering and Ajax; however, there was no consistent pattern between tritium release levels and Down Syndrome prevalence, chance could not be ruled out for the associations between Down Syndrome and tritium releases or ground-monitored concentrations, the association was detected in an analysis where multiple testing was done which may turn up significant associations by change, and maternal residence at birth and early in pregnancy needs to be verified. The association between Down Syndrome and low-level radiation remains indeterminate when existing evidence from epidemiological studies is summed. The estimated radiation exposure from the nuclear plant for residents of Pickering and Ajax is lower by a factor of 100 than the normal natural background radiation. Further study is recommended. (21 tabs., 29 figs., 5 maps, 37 refs.)

The interaction between popular economy, social movements and public policies: A case study of the waste pickers' movement

OpenAIRE

van Zeeland, Angelique J.W.M.

2014-01-01

This paper examines the challenges of expansion and sustainability of Social and Solidarity Economy (SSE). It focuses on the interaction between popular economy and SSE, and stresses the importance of collective action and public policies to enable the transition from the informal economy toward SSE. The main focus is on the waste pickers' movement. Experiences from Latin America, Asia and Africa show the possibilities of incorporating a significant contingent of informal waste pickers in sol...

System dynamics applied to closed loop supply chains of desktops and laptops in Brazil: A perspective for social inclusion of waste pickers.

Science.gov (United States)

Ghisolfi, Verônica; Diniz Chaves, Gisele de Lorena; Ribeiro Siman, Renato; Xavier, Lúcia Helena

2017-02-01

The structure of reverse logistics for waste electrical and electronic equipment (WEEE) is essential to minimize the impacts of their improper disposal. In this context, the Brazilian Solid Waste Policy (BSWP) was a regulatory milestone in Brazil, submitting WEEE to the mandatory implementation of reverse logistics systems, involving the integration of waste pickers on the shared responsibility for the life cycle of products. This article aims to measure the impact of such legal incentives and the bargaining power obtained by the volume of collected waste on the effective formalization of waste pickers. The proposed model evaluates the sustainability of supply chains in terms of the use of raw materials due to disposal fees, collection, recycling and return of some materials from desktops and laptops using system dynamics methodology. The results show that even in the absence of bargaining power, the formalization of waste pickers occurs due to legal incentives. It is important to ensure the waste pickers cooperatives access to a minimum amount, which requires a level of protection against unfair competition with companies. Regarding the optimal level of environmental policies, even though the formalization time is long, it is still not enough to guarantee the formalization of waste picker cooperatives, which is dependent on their bargaining power. Steel is the material with the largest decrease in acquisition rate of raw material. Copyright © 2016 Elsevier Ltd. All rights reserved.

AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1995

International Nuclear Information System (INIS)

1996-06-01

This report is the Atomic Energy Control Board (AECB) staff assessment of safety at the Pickering Nuclear Generating Station (PNGS-A and PNGS-B) for 1995. Our on-site Project Officers and Ottawa-based specialists monitored the stations throughout the year. In 1995, compliance with the Transportation Packaging of Radioactive Materials Regulations and the Cost Recovery Fees Regulations was satisfactory. The performance of the special safety systems was good. Releases of radioactive materials from the station were low and well below the legal limits for public safety. 10 tabs., 7 figs

Update of the Picker C9 irradiator control system of the gamma II room of the secondary laboratory of dosimetric calibration; Actualizacion del sistema de control del irradiador Picker C9 de la sala gamma II del laboratorio secundario de calibracion dosimetrica

Energy Technology Data Exchange (ETDEWEB)

Simon S, L. E.

2016-07-01

The Picker C9 irradiator is responsible for the calibration of different radiological equipment and the control system that maintains it in operation is designed in the graphical programming software LabVIEW (Laboratory Virtual Instrumentation Engineering Workbench), being its major advantages: the different types of communication, easy interconnection with other software and the recognition of different hardware devices, among others. Operation of the irradiator control system is performed with the NI-Usb-6008 (DAQ) data acquisition module of the National Instruments Company. The purpose of this work is to update the routines that make the Picker C9 control system of the gamma II room of the secondary laboratory of dosimetric calibration, using the graphic programming software LabVIEW, as well as to configure the new acquisition hardware of data that is implemented to control the Picker C9 irradiator system and ensure its operation. (Author)

Pickering Nuclear site wide groundwater monitoring system

International Nuclear Information System (INIS)

DeWilde, J.; Chin-Cheong, D.; Lledo, C.; Wootton, R.; Belanger, D.; Hansen, K.

2001-01-01

Ontario Power Generation Inc. (OPG) is continuing its efforts to understand the chemical and physical characteristics of the groundwater flow systems beneath the Pickering Nuclear Generating Station (PNGS). To this end, OPG constructed a site-wide Groundwater Monitoring System (GMS) at the PNGS to provide support to other ongoing environmental investigations and to provide a means to monitor current and future groundwater environmental issues. This paper will present the results of this work, including the development of a state-of-the-art data management system for storage and retrieval of environmental data for the site, which has applications for other power generation facilities. (author)

Pickering emulsions stabilized by whey protein nanoparticles prepared by thermal cross-linking

NARCIS (Netherlands)

Wu, Jiande; Shi, Mengxuan; Li, Wei; Zhao, Luhai; Wang, Ze; Yan, Xinzhong; Norde, Willem; Li, Yuan

2015-01-01

A Pickering (o/w) emulsion was formed and stabilized by whey protein isolate nanoparticles (WPI NPs). Those WPI NPs were prepared by thermal cross-linking of denatured WPI proteins within w/o emulsion droplets at 80. °C for 15. min. During heating of w/o emulsions containing 10% (w/v) WPI

Application of MCNPX 2.7.D for reactor core management at the research reactor BR2

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, Edgar

2011-01-01

The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK-CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full core model. The second method represents fully automatic whole core depletion and criticality calculations in the detailed 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared vs. the developed at the BR2 reactor department MCNPX & ORIGEN-S combined method. Testing of MCNPX 2.7.D on the criticality measurements at the BR2 reactor is presented. (author)

High-Surface-Area, Emulsion-Templated Carbon Foams by Activation of polyHIPEs Derived from Pickering Emulsions

Directory of Open Access Journals (Sweden)

Robert T. Woodward

2016-09-01

Full Text Available Carbon foams displaying hierarchical porosity and excellent surface areas of >1400 m2/g can be produced by the activation of macroporous poly(divinylbenzene. Poly(divinylbenzene was synthesized from the polymerization of the continuous, but minority, phase of a simple high internal phase Pickering emulsion. By the addition of KOH, chemical activation of the materials is induced during carbonization, producing Pickering-emulsion-templated carbon foams, or carboHIPEs, with tailorable macropore diameters and surface areas almost triple that of those previously reported. The retention of the customizable, macroporous open-cell structure of the poly(divinylbenzene precursor and the production of a large degree of microporosity during activation leads to tailorable carboHIPEs with excellent surface areas.

Waste picker livelihoods and inclusive neoliberal municipal solid waste management policies: The case of the La Chureca garbage dump site in Managua, Nicaragua.

Science.gov (United States)

Hartmann, Chris

2018-01-01

The modernization (i.e. mechanization, formalization, and capital intensification) and enclosure of municipal solid waste management (MSWM) systems threaten waste picker livelihoods. From 2009 to 2013, a major development project, embodying traditional neoliberal policies with inclusive social policies, transformed the Managua, Nicaragua, municipal solid waste site from an open-air dump where as many as 2,000 informal waste pickers toiled to a sanitary landfill. To investigate waste pickers' social and economic condition, including labor characteristics, household income, and poverty incidence, after the project's completion, 146 semi-structured survey questionnaires were administered to four communities adjacent to the landfill and 45 semi-structured interviews were completed with key stakeholders. Findings indicate that hundreds of waste pickers were displaced by the project, employment benefits from the project were unevenly distributed by neighborhood, and informal waste picking endures due to persistent impoverishment, thereby contributing to continued social and economic marginalization and environmental degradation. The findings highlight the limitations of inclusive neoliberal development efforts to transform MSWM in a low-income country. Copyright © 2017 Elsevier Ltd. All rights reserved.

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

Science.gov (United States)

Shen, Yongjun; Lei, Lecheng; Zhang, Xingwang; Ding, Jiandong

2014-11-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10-9 mol/L and 0.61 × 10-9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10-2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10-2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation pathways were

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

International Nuclear Information System (INIS)

Shen Yongjun; Ding Jiandong; Lei Lecheng; Zhang Xingwang

2014-01-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10 −9 mol/L and 0.61 × 10 −9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10 −2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10 −2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation

Multi-Unit Aspects of the Pickering Generating Station

Energy Technology Data Exchange (ETDEWEB)

Morison, W. G. [Atomic Energy of Canada Ltd, Sheridan Park, ON (Canada)

1968-04-15

The Pickering nuclear generating station is located on the north shore of Lake Ontario, about 20 miles east of the city of Toronto, Canada. The station has been planned and laid out on an eight-unit station, four units of which have now been authorized for construction. Each of these four units consists of a single heavy-water moderated and cooled CANDU-type reactor and auxiliaries coupled to a single tandem compound turbine generator with a net output of approximately 500 MW(e). The units are identical and are scheduled to come into operation at intervals of one year from 1970 to 1973. The station has been planned with central facilities for: administration maintenance laboratories, stores, change rooms, decontamination and waste management services. A common control centre, cooling water intake and discharge system, and spent fuel storage bay for four units has been arranged. A feature of the multi-unit station is a common containment system. Cost savings in building a number of identical units on the same site result from a single exclusion area, shared engineering costs, equipment purchase contracts for four identical components, and efficient use of construction plant. Operating cost savings are anticipated in the use of a common operating and maintenance staff and spare parts inventory. The plant has been arranged to minimize problems of operating, commissioning and constructing units at the same time on the same site. The layout and construction sequence have been arranged so that the first unit can be commissioned and operated with little or no interference from the construction forces working on succeeding units. During the construction phase barriers will be erected in the common control centre between operating control equipment and that being installed. Operations and construction personnel will enter the plant by separate routes and work in areas separated by physical barriers. (author)

The attached algae community near Pickering GS: III

International Nuclear Information System (INIS)

McKinley, S.R.

1982-01-01

The relationship between attached algae and macro-invertebrates in the nearshore zone of Lake Ontario was investigated in the vicinity of the Pickering 'A' NGS. Measures of faunal density, richness, evenness, and biomass were generally higher from areas which supported attached algae. Gammarus fasciatus, Cricotopus bicinctus, Dicrotendipes spp., Orthocladius obumbratus, Cladotanytarsus spp., Orthocladius spp., and Parakiefferiella spp., were significantly correlated with algal standing crop. All of the above dominant invertebrates ingested epiphytes associated with Cladophora glomerata. Attempts to explain the distribution of the zoobenthic assemblages using the physical/biological characteristics of the study area indicated algal cover, substrate size, wind velocity and water temperature were most important

Update of the Picker C9 irradiator control system of the gamma II room of the secondary laboratory of dosimetric calibration

International Nuclear Information System (INIS)

Simon S, L. E.

2016-01-01

The Picker C9 irradiator is responsible for the calibration of different radiological equipment and the control system that maintains it in operation is designed in the graphical programming software LabVIEW (Laboratory Virtual Instrumentation Engineering Workbench), being its major advantages: the different types of communication, easy interconnection with other software and the recognition of different hardware devices, among others. Operation of the irradiator control system is performed with the NI-Usb-6008 (DAQ) data acquisition module of the National Instruments Company. The purpose of this work is to update the routines that make the Picker C9 control system of the gamma II room of the secondary laboratory of dosimetric calibration, using the graphic programming software LabVIEW, as well as to configure the new acquisition hardware of data that is implemented to control the Picker C9 irradiator system and ensure its operation. (Author)

Effects of Occupational Exposure on the Health of Rag Pickers Due to Fungal Contamination at Waste Dumping Sites in Gwalipor (India

Directory of Open Access Journals (Sweden)

Harandra K. Sharma

2017-02-01

Full Text Available We investigated fungal contamination near different waste dumping sites and assessed the health risk factors of rag pickers associated with collection of waste in Gwalior during the year 2014-15. Petri plates were exposed at waste dumping sites and were transferred to the laboratory, analysis and identification was mainly carried out by culturing the fungal colonies by following standard procedures. A pretested questionnaire was used to evaluate the health problems among the rag pickers. Results indicated that all the dumping sites are contaminated with different types of fungal pathogens like Alternaria alternate, Aspergillus flavus, A. fumigates, A. niger, Cladosporium, Fusarium, Mucor, Penicillium and Rhizopus. Our study reported higher incidence of musculoskeletal and respiratory diseases among rag pickers. There is also strong need for carrying out similar assessment studies for other cities too. This will entail generation of more precise site specific information regarding fungal species and associated health risk factor.

Effects of Occupational Exposure on the Health of Rag Pickers Due to Fungal Contamination at Waste Dumping Sites in Gwalior (India

Directory of Open Access Journals (Sweden)

Harandra K. Sharma

2017-02-01

Full Text Available We investigated fungal contamination near different waste dumping sites and assessed the health risk factors of rag pickers associated with collection of waste in Gwalior during the year 2014-15. Petri plates were exposed at waste dumping sites and were transferred to the laboratory, analysis and identification was mainly carried out by culturing the fungal colonies by following standard procedures. A pretested questionnaire was used to evaluate the health problems among the rag pickers. Results indicated that all the dumping sites are contaminated with different types of fungal pathogens like Alternaria alternate, Aspergillus flavus, A. fumigates, A. niger, Cladosporium, Fusarium, Mucor, Penicillium and Rhizopus. Our study reported higher incidence of musculoskeletal and respiratory diseases among rag pickers. There is also strong need for carrying out similar assessment studies for other cities too. This will entail generation of more precise site specific information regarding fungal species and associated health risk factor.

Hydroxyapatite-armored poly(ε-caprolactone) microspheres and hydroxyapatite microcapsules fabricated via a Pickering emulsion route.

Science.gov (United States)

Fujii, Syuji; Okada, Masahiro; Nishimura, Taiki; Maeda, Hayata; Sugimoto, Tatsuya; Hamasaki, Hiroyuki; Furuzono, Tsutomu; Nakamura, Yoshinobu

2012-05-15

Hydroxyapatite (HAp) nanoparticle-armored poly(ε-caprolactone) (PCL) microspheres were fabricated via a "Pickering-type" emulsion solvent evaporation method in the absence of any molecular surfactants. It was clarified that the interaction between carbonyl/carboxylic acid groups of PCL and the HAp nanoparticles at an oil-water interface played a crucial role in the preparation of the stable Pickering-type emulsions and the HAp nanoparticle-armored microspheres. The HAp nanoparticle-armored PCL microspheres were characterized in terms of size, size distribution, morphology, and chemical compositions using scanning electron microscopy, laser diffraction, energy dispersive X-ray microanalysis, and thermogravimetric analysis. The presence of HAp nanoparticles at the surface of the microspheres was confirmed by scanning electron microscopy and energy dispersive X-ray microanalysis. Pyrolysis of the PCL cores led to the formation of the corresponding HAp hollow microcapsules. Copyright © 2012 Elsevier Inc. All rights reserved.

Pickering emulsion: A novel template for microencapsulated phase change materials with polymer–silica hybrid shell

International Nuclear Information System (INIS)

Yin, Dezhong; Ma, Li; Liu, Jinjie; Zhang, Qiuyu

2014-01-01

MePCMs (microencapsulated phase change materials) with covalently bonded SiO 2 /polymer hybrid as shell were fabricated via Pickering emulsion polymerization stabilized solely by organically-modified SiO 2 particles. Morphology and core–shell structure of these microcapsules were observed by scanning electron microscopy (SEM). Thermal properties of microencapsulated 1-dodecanol were determined using DSC (differential scanning calorimetry) and TGA (thermal gravimetric analysis). The results indicate that mass ratio of St (styrene)/DVB (divinylbenzene)/dodecanol has great effect on the morphology, inner structure, microencapsulation efficiency and durability of resultant MePCMs. When ratio of St/DVB/dodecanol was 5/1/12, dodecanol content of as much as 62.8% is obtained and the utility efficiency of dodecanol reaches 94.2%. The prepared MePCMs present good durability and thermal reliability. 2.2% of core material leached away the microcapsule after suspended in water for 10 days and 5.8% of core material leached after 2000 accelerated thermal cycling. Our study demonstrated that Pickering emulsion polymerization is a simple and robust method for the preparation of MePCMs with polymer–inorganic hybrids as shell. - Highlights: • We fabricated MePCM via surfactant-free Pickering emulsion polymerization. • The shell of MePCM was composed of PS/SiO 2 organic–inorganic hybrids. • The phase change enthalpy of MePCM is 125.0 J g −1 and the utility efficiency of 1-dodecanol reached 94.2%. • Only 2.2% and 5.8% of core material lost after durability test and 2000 accelerated thermal cycling respectively

Pickering irradiated fuel transfer conveyor isolation

Energy Technology Data Exchange (ETDEWEB)

Koivisto, D J; Eijsermans, L J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1997-12-31

Pickering A NGS has been in operation for 25 years and is one of the longest in service CANDU stations. Some underwater fuel handling equipment, notably the conveyor stops, have been without maintenance throughout that time. This paper describes the concept of a conveyor isolation system that permits draining of a single or multiple elevator columns and also the early stages of a development program for the elastomeric sealing element. The prototype seal element has been proven in lab tests to be capable of limiting leakage to 0.5 IGPM (imperial gallons per minute) at the design pressure of 6.5 psi. The design of a sealing element is particularly interesting because the conveyor tube is a square cross-section which contains an additional obstruction , a conveyor drive cable. A seal delivery, actuating and positioning system has been conceptually laid out and the design is proceeding, with projected implementation in 1998. (author). 8 figs.

Pickering irradiated fuel transfer conveyor isolation

International Nuclear Information System (INIS)

Koivisto, D.J.; Eijsermans, L.J.

1996-01-01

Pickering A NGS has been in operation for 25 years and is one of the longest in service CANDU stations. Some underwater fuel handling equipment, notably the conveyor stops, have been without maintenance throughout that time. This paper describes the concept of a conveyor isolation system that permits draining of a single or multiple elevator columns and also the early stages of a development program for the elastomeric sealing element. The prototype seal element has been proven in lab tests to be capable of limiting leakage to 0.5 IGPM (imperial gallons per minute) at the design pressure of 6.5 psi. The design of a sealing element is particularly interesting because the conveyor tube is a square cross-section which contains an additional obstruction , a conveyor drive cable. A seal delivery, actuating and positioning system has been conceptually laid out and the design is proceeding, with projected implementation in 1998. (author). 8 figs

Uplifting : mobile cranes and picker trucks get higher, stronger, and safer

Energy Technology Data Exchange (ETDEWEB)

Budd, G.

2008-10-15

New crane and picker truck equipment designed for use in the oil and gas industry was discussed in this article. The longest crane in the North America is due to arrive in Calgary soon. Mobile cranes are often used for maintenance, construction, and shutdowns at natural gas plants, refineries, and oil batteries. Telescopic cranes are also used to service pump jacks and lift rolls of coiled tubing into place on drilling rigs. While cranes carry more weight, picker trucks are more mobile and flexible. Lattice boom crawler cranes were designed to pick up loads and carry them to different positions. Telescopic cranes must remain stationary, and careful planning is needed to ensure that they are used efficiently. Modern telescopic cranes have hydraulically powered booms. Advanced steel alloy technology is used to produce lightweight booms equipped with automated pin-locking mechanisms. The largest telescopic crane in North America will be used by an oil and gas operator in Alberta. The crane will have a combined boom and lattice jib height of 226 meters, and its maximum lifting capacity is 1200 tonnes. Gantry cranes are also increasingly being used by oil and gas operators due to their ability to straddle loads, thereby reducing the risk of tipping. It was concluded that gantry cranes are particularly suitable for rougher terrains. 4 figs.

Online control loop tuning in Pickering Nuclear Generating Stations

International Nuclear Information System (INIS)

Yu, K.X.; Harrington, S.

2008-01-01

Most analog controllers in the Pickering B Nuclear Generating Stations adopted PID control scheme. In replacing the analog controllers with digital controllers, the PID control strategies, including the original tuning parameters were retained. The replacement strategy resulted in minimum effort on control loop tuning. In a few cases, however, it was found during commissioning that control loop tuning was required as a result of poor control loop performance, typically due to slow response and controlled process oscillation. Several factors are accounted for the necessities of control loop re-tuning. Our experience in commissioning the digital controllers showed that online control tuning posted some challenges in nuclear power plant. (author)

Networks of recyclable material waste-picker's cooperatives: an alternative for the solid waste management in the city of Rio de Janeiro.

Science.gov (United States)

Tirado-Soto, Magda Martina; Zamberlan, Fabio Luiz

2013-04-01

The objective of this study is to discuss the role of networks formed of waste-picker cooperatives in ameliorating problems of final disposal of solid waste in the city of Rio de Janeiro, since the city's main landfill will soon have to close because of exhausted capacity. However, it is estimated that in the city of Rio de Janeiro there are around five thousand waste-pickers working in poor conditions, with lack of physical infrastructure and training, but contributing significantly by diverting solid waste from landfills. According to the Sustainable Development Indicators (IBGE, 2010a,b) in Brazil, recycling rates hover between 45% and 55%. In the municipality of Rio de Janeiro, only 1% of the waste produced is collected selectively by the government (COMLURB, 2010), demonstrating that recycling is mainly performed by waste-pickers. Furthermore, since the recycling market is an oligopsony that requires economies of scale to negotiate directly with industries, the idea of working in networks of cooperatives meets the demands for joint marketing of recyclable materials. Thus, this work presents a method for creating and structuring a network of recycling cooperatives, with prior training for working in networks, so that the expected synergies and joint efforts can lead to concrete results. We intend to demonstrate that it is first essential to strengthen the waste-pickers' cooperatives in terms of infrastructure, governance and training so that solid waste management can be environmentally, socially and economically sustainable in the city of Rio de Janeiro. Copyright © 2012 Elsevier Ltd. All rights reserved.

Arresting relaxation in Pickering Emulsions

Science.gov (United States)

Atherton, Tim; Burke, Chris

2015-03-01

Pickering emulsions consist of droplets of one fluid dispersed in a host fluid and stabilized by colloidal particles absorbed at the fluid-fluid interface. Everyday materials such as crude oil and food products like salad dressing are examples of these materials. Particles can stabilize non spherical droplet shapes in these emulsions through the following sequence: first, an isolated droplet is deformed, e.g. by an electric field, increasing the surface area above the equilibrium value; additional particles are then adsorbed to the interface reducing the surface tension. The droplet is then allowed to relax toward a sphere. If more particles were adsorbed than can be accommodated by the surface area of the spherical ground state, relaxation of the droplet is arrested at some non-spherical shape. Because the energetic cost of removing adsorbed colloids exceeds the interfacial driving force, these configurations can remain stable over long timescales. In this presentation, we present a computational study of the ordering present in anisotropic droplets produced through the mechanism of arrested relaxation and discuss the interplay between the geometry of the droplet, the dynamical process that produced it, and the structure of the defects observed.

Histidine-functionalized carbon-based dot-Zinc(II) nanoparticles as a novel stabilizer for Pickering emulsion synthesis of polystyrene microspheres.

Science.gov (United States)

Ruiyi, Li; Zaijun, Li; Junkang, Liu

2017-05-01

Carbon-based dots (CDs) are nanoparticles with size-dependent optical and electronic properties that have been widely applied in energy-efficient displays and lighting, photovoltaic devices and biological markers. However, conventional CDs are difficult to be used as ideal stabilizer for Pickering emulsion due to its irrational amphiphilic structure. The study designed and synthesized a new histidine-functionalized carbon dot-Zinc(II) nanoparticles, which is termed as His-CD-Zn. The His-CD was made via one-step hydrothermal treatment of histidine and maleic acid. The His-CD reacted with Zn 2+ to form His-CD-Zn. The as-prepared His-CD-Zn was used as a solid particle surfactant for stabilizing styrene-in-water emulsion. The Pickering emulsion exhibits high stability and sensitive pH-switching behaviour. The introduction of S 2 O 8 2- triggers the emulsion polymerization of styrene. The resulted polystyrene microsphere was well coated with His-CDs on the surface. It was successfully used as an ideal adsorbent for removal of heavy metallic ions from water with high adsorption capacity. The study also provides a prominent approach for fabrication of amphiphilic carbon-based nanoparticles for stabilizing Pickering emulsion. Copyright © 2017 Elsevier Inc. All rights reserved.

Pickering G.S. boiler repair: an example of planned maintenance

International Nuclear Information System (INIS)

Dalrymple, D.G.

1976-04-01

The first application of boiler repair tools and procedures is estimated to have yielded a four-fold return on the development investment. The need to develop such technology is a result of the environment in which boiler repairs must be made. As nuclear technology evolves and plants and components get bigger, equipment will increasingly have to be repaired in situ with minimum plant downtime and minimum exposure of repair personnel to radiation. This lecture traces development of the Pickering A boiler repair capability which is seen as an example of how utility and contractor should interact to anticipate and meet maintenance requirements. (author)

Pickering NGS emergency water supply system emergency start flow simulation and experiment

Energy Technology Data Exchange (ETDEWEB)

Davidge, E.; Misra, A. [Ontario Power Generation Inc., Nuclear Safety Analysis & Technology Department, Toronto, Ontario (Canada)

2012-07-01

A proposed modification to the OPG Pickering Nuclear Generation Station Emergency Water Supply (EWS) system was analyzed using the Industry Standard Toolset code GOTHIC to determine the acceptability of the proposed system configuration during pump start-up. The new configuration of the system included a vertical dead-ended pipe, initially filled with air. The simulation demonstrated that no significant water hammer effects were predicted and tests performed with the new configuration confirmed the analysis results. (author)

AECB staff annual report on Pickering NGS for the year 1989

International Nuclear Information System (INIS)

1990-06-01

This report presents a review of major licensing issues and the operational performance of Pickering NGS-A (Units 1-4) and Pickering NGS-B (Units 5-8) by the staff of the Atomic Energy Control Board (AECB) during 1989. Operations are monitored to ensure compliance with licensing requirements. This report is limited to those aspects that AECB staff consider to have particular safety significance. The number of infractions of the operating licence and the AECB Regulations doubled in 1989 compared to 1988. Three workers were exposed to radiation doses in excess of the regulatory limits. The AECB also found inadequate procedural compliance and an unacceptable level of housekeeping. Performance also requires improvement in response to AECB Health Physics appraisals; surveillance of systems by the Technical Section; chemical control performance; response to outstanding AECB action items; availability of special safety systems; operating memos, jumper records, call-ups and deficiency reports in effect; and fire fighting capability. Ontario Hydro has initiated a number of programs that are expected to result in improvements in these areas: an in-service station quality improvement plan; a program to improve and give assurance of compliance with the AECB Regulations, the operating licenses and the Operating Policies and Principles; a housekeeping and material condition improvement plan; and an action plan undertaken following radiation over-exposures. Given adequate resources and management support these programs should result in a noticeable improvement in station performance in 1990

Phytotoxicology section investigation in the vicinity of the Bruce Nuclear Power Development, the Pickering Nuclear Generating Station and the Darlington Nuclear Generating Station, in October, 1989

International Nuclear Information System (INIS)

1991-02-01

The Phytotoxicology Section, Air Resources Branch is a participant in the Pickering and Bruce Nuclear Contingency Plans. The Phytotoxicology Emergency Response Team is responsible for collecting vegetation samples in the event of a nuclear emergency at any of the nuclear generating stations in the province. As part of its responsibility the Phytotoxicology Section collects samples around the nuclear generating stations for comparison purposes in the event of an emergency. Because of the limited frequency of sampling, the data from the surveys are not intended to be used as part of a regulatory monitoring program. These data represent an effort by the MOE to begin to establish a data base of tritium concentrations in vegetation. The Phytotoxicology Section has carried out seven surveys in the vicinity of Ontario Hydro nuclear generating stations since 1981. Surveys were conducted for tritium in snow in the vicinity of Bruce Nuclear Power Development (BNPD), February, 1981; tritium in cell-free water of white ash in the vicinity of BNPD, September, 1981; tritium in snow in the vicinity of BNPD, March, 1982; tritium in tree sap in the vicinity of BNPD, April, 1982; tritium in tree sap in the vicinity of BNPD, April, 1984, tritium in the cell-free water of white ash in the vicinity of BNPD, September, 1985; and, tritium in cell-free water of grass in the vicinity of Pickering Nuclear Generation Station (PNGS), October 1986. In all cases a pattern of decreasing tritium levels with increasing distance from the stations was observed. In October, 1989, assessment surveys were conducted around Bruce Nuclear Power Development, the Pickering Nuclear Generating Station and the new Darlington Nuclear Generating Station (DNGS). The purpose of these surveys was to provide baseline data for tritium in cell-free water of grass at all three locations at the same time of year. As none of the reactor units at DNGS had been brought on line at the time of the survey, this data was to be

Field testing of behavioral barriers for cooling water intake structures -test site 1 - Pickering Nuclear Generating Station

International Nuclear Information System (INIS)

Patrick, P.H.; McKinley, R.S.; Micheletti, W.C.

1988-01-01

A multi-year research program was developed by the Electric Power Research Institute to evaluate the effectiveness of selected behavioral systems for fish exclusion at sites representative of different aquatic environments. The first test site was the Pickering Nuclear Generating Station (NGS) located on Lake Ontario which represented the Great Lakes environment. A single pneumatic popper, a low frequency, high amplitude sound deterrent, was found to effectively exclude adult alewife, the principal species impinged at Pickering NGS. An air bubble curtain, used either alone or combined with strobe lights, was not a consistent deterrent. Effectiveness of air bubbles was only enhanced when used in association with a popper. Strobe lights were the least effective of the three devices tested. Operation of all three devices together did not surpass the effectiveness of the popper when used alone. Sound deterrents show promise for fish exclusion at generating stations located on the Great Lakes

Viscosity of the oil-in-water Pickering emulsion stabilized by surfactant-polymer and nanoparticle-surfactant-polymer system

Science.gov (United States)

Sharma, Tushar; Kumar, G. Suresh; Chon, Bo Hyun; Sangwai, Jitendra S.

2014-11-01

Information on the viscosity of Pickering emulsion is required for their successful application in upstream oil and gas industry to understand their stability at extreme environment. In this work, a novel formulation of oil-in-water (o/w) Pickering emulsion stabilized using nanoparticle-surfactant-polymer (polyacrylamide) system as formulated in our earlier work (Sharma et al., Journal of Industrial and Engineering Chemistry, 2014) is investigated for rheological stability at high pressure and high temperature (HPHT) conditions using a controlled-strain rheometer. The nanoparticle (SiO2 and clay) concentration is varied from 1.0 to 5.0 wt%. The results are compared with the rheological behavior of simple o/w emulsion stabilized by surfactant-polymer system. Both the emulsions exhibit non-Newtonian shear thinning behavior. A positive shift in this behavior is observed for surfactant-polymer stabilized emulsion at high pressure conditions. Yield stress is observed to increase with pressure for surfactant-polymer emulsion. In addition, increase in temperature has an adverse effect on the viscosity of emulsion stabilized by surfactant-polymer system. In case of nanoparticle-surfactant-polymer stabilized o/w emulsion system, the viscosity and yield stress are predominantly constant for varying pressure and temperature conditions. The viscosity data for both o/w emulsion systems are fitted by the Herschel-Bulkley model and found to be satisfactory. In general, the study indicates that the Pickering emulsion stabilized by nanoparticle-surfactant-polymer system shows improved and stable rheological properties as compared to conventional emulsion stabilized by surfactant-polymer system indicating their successful application for HPHT environment in upstream oil and gas industry.

Quinoa starch granules as stabilizing particles for production of Pickering emulsions.

Science.gov (United States)

Rayner, Marilyn; Sjöö, Malin; Timgren, Anna; Dejmek, Petr

2012-01-01

Intact starch granules isolated from quinoa (Chenopodium quinoa Willd.) were used to stabilize emulsion drops in so-called Pickering emulsions. Miglyol 812 was used as dispersed phase and a phosphate buffer (pH7) with different salt (NaCl) concentrations was used as the continuous phase. The starch granules were hydrophobically modified to different degrees by octenyl succinic anhydride (OSA) or by dry heat treatment at 120 degrees C in order to study the effect on the resulting emulsion drop size. The degree of OSA-modification had a low to moderate impact on drop size. The highest level of modification (4.66%) showed the largest mean drop size, and lowest amount of free starch, which could be an effect of a higher degree of aggregation of the starch granules and, thereby, also the emulsion drops stabilized by them. The heat treated starch granules had a poor stabilizing ability and only the starch heated for the longest time (150 min at 120 degrees C) had a better emulsifying capacity than the un-modified native starch granules. The effect of salt concentration was rather limited. However, an increased concentration of salt slightly increased the mean drop size and the elastic modulus.

Waste Picker Organizations and Their Contribution to the Circular Economy: Two Case Studies from a Global South Perspective

Directory of Open Access Journals (Sweden)

Jutta Gutberlet

2017-09-01

Full Text Available The discussion on the circular economy (CE has attracted a rising interest within global policy and business as a way of increasing the sustainability of production and consumption. Yet the literature mostly portrays a Global North perspective. There is a diverse spectrum of community-based organizations playing important roles in resource recovery and transformation, particularly, but not only, in Global South countries, providing innovative examples for grassroots involvement in waste management and in the CE. This article proposes to add a Southern lens, situated in the context of waste picker organizations, to the concept of CE. The discursive framework in this article couples ecological economy (EE with social/solidarity economy (SSE, focusing not only on environmental sustainability but also on social, economic, political and cultural dimensions involved in production, consumption and discard. We acknowledge that grassroots movements contribute to policy making and improve urban waste management systems. The paper outlines two empirical studies (Argentina, Brazil that illustrate how waste picker organizations perform selective waste collection services, engage with municipalities and industries, and practice the CE. The research reveals that social and political facets need to be added to the debate about the CE, linking environmental management and policy with community development and recognizing waste pickers as protagonists in the CE. Our findings emphasize a need for a change of persisting inequalities in public policy by recognizing the importance of popular waste management praxis and knowledge, ultimately redefining the CE.

Expanding worldwide urban solid waste recycling: The Brazilian social technology in waste pickers inclusion.

Science.gov (United States)

Rutkowski, Jacqueline E; Rutkowski, Emília W

2015-12-01

'If an integrated urban waste management system includes the informal recycling sector (IRS), there is a good chance that more solid waste is recycled' is common sense. However, informal integration brings additional social, environmental, and economic benefits, such as reduction of operational costs and environmental impacts of landfilling. Brazil is a global best practice example in terms of waste picker inclusion, and has received international recognition for its recycling levels. In addition to analysing the results of inclusive recycling approaches, this article evaluates a selection of the best Brazilian inclusive recycling practices and summaries and presents the resulting knowledge. The objective is to identify processes that enable the replication of the inclusion of the informal recycling sector model as part of municipal solid waste management. Qualitative and quantitative data have been collected in 25 Brazilian cities that have contracted waste pickers co-operatives for door-to-door selective collection of recyclables. Field data was collected in action research projects that worked with waste pickers co-operatives between 2006 and 2013. The Brazilian informal recycling sector integration model improves municipal solid waste recycling indicators: it shows an increase in the net tonness recycled, from 140 to 208 t month(-1), at a much lower cost per tonne than conventional selective collection systems. Inclusive systems show costs of US$35 per tonne of recyclables collected, well below the national average of US$195.26. This inclusive model improves the quality of collected material and the efficiency of municipal selective collection. It also diminishes the negative impacts of informal recycling, by reducing child labour, and by improving the conditions of work, occupational health and safety, and uncontrolled pollution. Although treating the Brazilian experience as a blueprint for transfer of experience in every case is unrealistic, the results

Colloidal formulations for probiotics delivery and Pickering systems

DEFF Research Database (Denmark)

Yücel Falco, Cigdem

countries. One emerging functional food area is the efficient delivery of health-promoting probiotics. Although much progress has already been made in the development and understanding of novel microencapsulation systems, maintaining viability during gastric passage and being effective at the target site...... is still an issue for probiotics. On the other hand, one of the foremost challenges in the production of physically stable foods during the defined shelf life is the identification of new food-grade ingredients. In this context, the replacement of classical emulsifiers with solid particles is one...... of the advancing food research areas, though the number of food-grade solid particles investigated is still insufficient. Edible probiotic strains can potentially be valorised as particles similar to micron-sized fat particles in Pickering systems such as ice cream due to their low calories and their availability...

Development and applications of reactor noise analysis at Ontario Hydro's CANDU reactors

International Nuclear Information System (INIS)

Gloeckler, O.; Tulett, M.V.

1995-01-01

In 1992 a program was initiated to establish reactor noise analysis as a practical tool for plant performance monitoring and system diagnostics in Ontario Hydro's CANDU reactors. Since then, various CANDU-specific noise analysis applications have been developed and validated. The noise-based statistical techniques are being successfully applied as powerful troubleshooting and diagnostic tools to a wide variety of actual operational I and C problems. The dynamic characteristics of critical plant components, instrumentation and processes are monitored on a regular basis. Recent applications of noise analysis include (1) validating the dynamics of in-core flux detectors (ICFDS) and ion chambers, (2) estimating the prompt fraction ICFDs in noise measurements at full power and in power rundown tests, (3) identifying the cause of excessive signal fluctuations in certain flux detectors, (4) validating the dynamic coupling between liquid zone control signals, (5) detecting and monitoring mechanical vibrations of detector tubes induced by moderator flow, (6) estimating the dynamics and response time of RTD (Resistance Temperature Detector) temperature signals, (7) isolating the cause of RTD signal anomalies, (8) investigating the source of abnormal flow signal behaviour, (9) estimating the overall response time of flow and pressure signals, (10) detecting coolant boiling in fully instrumented fuel channels, (11) monitoring moderator circulation via temperature noise, and (12) predicting the performance of shut-off rods. Some of these applications are performed on an as-needed basis. The noise analysis program, in the Pickering-B station alone, has saved Ontario Hydro millions of dollars during its first three years. The results of the noise analysis program have been also reviewed by the regulator (Atomic Energy Control Board of Canada) with favorable results. The AECB have expressed interest in Ontario Hydro further exploiting the use of noise analysis technology. (author

Reverse logistics network for municipal solid waste management: The inclusion of waste pickers as a Brazilian legal requirement

International Nuclear Information System (INIS)

Ferri, Giovane Lopes; Diniz Chaves, Gisele de Lorena; Ribeiro, Glaydston Mattos

2015-01-01

Highlights: • We propose a reverse logistics network for MSW involving waste pickers. • A generic facility location mathematical model was validated in a Brazilian city. • The results enable to predict the capacity for screening and storage centres (SSC). • We minimise the costs for transporting MSW with screening and storage centres. • The use of SSC can be a potential source of revenue and a better use of MSW. - Abstract: This study proposes a reverse logistics network involved in the management of municipal solid waste (MSW) to solve the challenge of economically managing these wastes considering the recent legal requirements of the Brazilian Waste Management Policy. The feasibility of the allocation of MSW material recovery facilities (MRF) as intermediate points between the generators of these wastes and the options for reuse and disposal was evaluated, as well as the participation of associations and cooperatives of waste pickers. This network was mathematically modelled and validated through a scenario analysis of the municipality of São Mateus, which makes the location model more complete and applicable in practice. The mathematical model allows the determination of the number of facilities required for the reverse logistics network, their location, capacities, and product flows between these facilities. The fixed costs of installation and operation of the proposed MRF were balanced with the reduction of transport costs, allowing the inclusion of waste pickers to the reverse logistics network. The main contribution of this study lies in the proposition of a reverse logistics network for MSW simultaneously involving legal, environmental, economic and social criteria, which is a very complex goal. This study can guide practices in other countries that have realities similar to those in Brazil of accelerated urbanisation without adequate planning for solid waste management, added to the strong presence of waste pickers that, through the

Reverse logistics network for municipal solid waste management: The inclusion of waste pickers as a Brazilian legal requirement

Energy Technology Data Exchange (ETDEWEB)

Ferri, Giovane Lopes, E-mail: giovane.ferri@aluno.ufes.br [Department of Engineering and Technology, Federal University of Espírito Santo – UFES, Rodovia BR 101 Norte, Km 60, Bairro Litorâneo, São Mateus, ES, 29.932-540 (Brazil); Diniz Chaves, Gisele de Lorena, E-mail: gisele.chaves@ufes.br [Department of Engineering and Technology, Federal University of Espírito Santo – UFES, Rodovia BR 101 Norte, Km 60, Bairro Litorâneo, São Mateus, ES, 29.932-540 (Brazil); Ribeiro, Glaydston Mattos, E-mail: glaydston@pet.coppe.ufrj.br [Transportation Engineering Programme, Federal University of Rio de Janeiro – UFRJ, Centro de Tecnologia, Bloco H, Sala 106, Cidade Universitária, Rio de Janeiro, 21949-900 (Brazil)

2015-06-15

Highlights: • We propose a reverse logistics network for MSW involving waste pickers. • A generic facility location mathematical model was validated in a Brazilian city. • The results enable to predict the capacity for screening and storage centres (SSC). • We minimise the costs for transporting MSW with screening and storage centres. • The use of SSC can be a potential source of revenue and a better use of MSW. - Abstract: This study proposes a reverse logistics network involved in the management of municipal solid waste (MSW) to solve the challenge of economically managing these wastes considering the recent legal requirements of the Brazilian Waste Management Policy. The feasibility of the allocation of MSW material recovery facilities (MRF) as intermediate points between the generators of these wastes and the options for reuse and disposal was evaluated, as well as the participation of associations and cooperatives of waste pickers. This network was mathematically modelled and validated through a scenario analysis of the municipality of São Mateus, which makes the location model more complete and applicable in practice. The mathematical model allows the determination of the number of facilities required for the reverse logistics network, their location, capacities, and product flows between these facilities. The fixed costs of installation and operation of the proposed MRF were balanced with the reduction of transport costs, allowing the inclusion of waste pickers to the reverse logistics network. The main contribution of this study lies in the proposition of a reverse logistics network for MSW simultaneously involving legal, environmental, economic and social criteria, which is a very complex goal. This study can guide practices in other countries that have realities similar to those in Brazil of accelerated urbanisation without adequate planning for solid waste management, added to the strong presence of waste pickers that, through the

Intestinal parasitism among waste pickers in Mato Grosso do Sul, Midwest Brazil

Directory of Open Access Journals (Sweden)

Minoru German Higa Júnior

2017-12-01

Full Text Available ABSTRACT The purpose of this study was to estimate the prevalence of intestinal parasites in both cooperative-affiliated and independent waste pickers operating at the municipal sanitary landfill in Campo Grande, Mato Grosso do Sul, Brazil, and associate these findings with hemoglobin, eosinophils, vitamin A and C levels and interleukin 5 and 10 (IL-5 and IL-10 production. Biological samples were collected, in addition to clinical, epidemiological, and sociodemographic data. Stool analyzes were based on sedimentation by centrifugation and on spontaneous sedimentation. High-performance liquid chromatography was used to determine vitamin A and C levels. ELISA was employed to quantify interleukins. Intestinal parasites were found in 29 of the 66 subjects assessed (43.9%. Endolimax nana (22.7%, Entamoeba coli (21.1%, Giardia lamblia (6.1%, Entamoeba histolytica/E. dispar (4.5%, and Ascaris lumbricoides (4.5% were the most prevalent species. Pathogenic parasites were detected in 11 individuals (16.7%. Hypovitaminoses A and C were detected in 19.6% (13/66 and 98.4% (65/66 of subjects, respectively. IL-5 and IL-10 production was observed in 21 (31.8% and 32 (48.4% subjects, respectively. Infection with pathogenic intestinal parasites was not a cause of vitamin A and C deficiency or IL-5 and IL-10 production among these workers.

Pickering emulsions stabilized by biodegradable block copolymer micelles for controlled topical drug delivery.

Science.gov (United States)

Laredj-Bourezg, Faiza; Bolzinger, Marie-Alexandrine; Pelletier, Jocelyne; Chevalier, Yves

2017-10-05

Surfactant-free biocompatible and biodegradable Pickering emulsions were investigated as vehicles for skin delivery of hydrophobic drugs. O/w emulsions of medium-chain triglyceride (MCT) oil droplets loaded with all-trans retinol as a model hydrophobic drug were stabilized by block copolymer nanoparticles: either poly(lactide)-block-poly(ethylene glycol) (PLA-b-PEG) or poly(caprolactone)-block-poly(ethylene glycol) (PCL-b-PEG). Those innovative emulsions were prepared using two different processes allowing drug loading either inside oil droplets or inside both oil droplets and non-adsorbed block copolymer nanoparticles. Skin absorption of retinol was investigated in vitro on pig skin biopsies using the Franz cell method. Supplementary experiments by confocal fluorescence microscopy allowed the visualization of skin absorption of the Nile Red dye on histological sections. Retinol and Nile Red absorption experiments showed the large accumulation of hydrophobic drugs in the stratum corneum for the Pickering emulsions compared to the surfactant-based emulsion and an oil solution. Loading drug inside both oil droplets and block copolymer nanoparticles enhanced again skin absorption of drugs, which was ascribed to the supplementary contribution of free block copolymer nanoparticles loaded with drug. Such effect allowed tuning drug delivery to skin over a wide range by means of a suitable selection of either the formulation or the drug loading process. Copyright © 2017 Elsevier B.V. All rights reserved.

Diagrams about RA reactor operation, Annex 7; Prilog 7 - Rad reaktora RA po godinama - dijagrami

Energy Technology Data Exchange (ETDEWEB)

Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1976-01-15

This Annex 7 includes diagrams about RA reactor operation (MWh) from 1960-1975; mean values of reactor power per day and month in 1975, and percent of utilisation of experimental space during 1975. [Serbo-Croat] Ovaj prilog sadrzi dijagrame o radu reaktora (MWh) po godinama (1960-1975); srednje vrednosti dnevne snage reaktora u 1975. godini; rad reakrora (MWh) po mesecima u 1975; procenat iskoriscenja eksperimentalnog prostora u 1975. godini.

Selective removal of erythromycin by magnetic imprinted polymers synthesized from chitosan-stabilized Pickering emulsion.

Science.gov (United States)

Ou, Hongxiang; Chen, Qunhui; Pan, Jianming; Zhang, Yunlei; Huang, Yong; Qi, Xueyong

2015-05-30

Magnetic imprinted polymers (MIPs) were synthesized by Pickering emulsion polymerization and used to adsorb erythromycin (ERY) from aqueous solution. The oil-in-water Pickering emulsion was stabilized by chitosan nanoparticles with hydrophobic Fe3O4 nanoparticles as magnetic carrier. The imprinting system was fabricated by radical polymerization with functional and crosslinked monomer in the oil phase. Batches of static and dynamic adsorption experiments were conducted to analyze the adsorption performance on ERY. Isotherm data of MIPs well fitted the Freundlich model (from 15 °C to 35 °C), which indicated heterogeneous adsorption for ERY. The ERY adsorption capacity of MIPs was about 52.32 μmol/g at 15 °C. The adsorption kinetics was well described by the pseudo-first-order model, which suggested that physical interactions were primarily responsible for ERY adsorption. The Thomas model used in the fixed-bed adsorption design provided a better fit to the experimental data. Meanwhile, ERY exhibited higher affinity during adsorption on the MIPs compared with the adsorption capacity of azithromycin and chloramphenicol. The MIPs also exhibited excellent regeneration capacity with only about 5.04% adsorption efficiency loss in at least three repeated adsorption-desorption cycles. Copyright © 2015 Elsevier B.V. All rights reserved.

Investigation of tritium in groundwater at Pickering NGS

International Nuclear Information System (INIS)

DeWilde, J.; Yu, L.; Belanger, D.; Wootton, R.; Hansen, K.; McGurk, E.; Teare, A.

2001-01-01

Ontario Power Generation Inc. (OPG) investigated tritium in groundwater at the Pickering Nuclear Generating Station (PNGS). The objectives of the study were to evaluate and define the extent of radio-nuclides, primarily tritium, in groundwater, investigate the causes or sources of contamination, determine impacts on the natural environment, and provide recommendations to prevent future discharges. This paper provides an overview of the investigations conducted in 1999 and 2000 to identify the extent of the tritium beneath the site and the potential sources of tritium released to the groundwater. The investigation and findings are summarized with a focus on unique aspects of the investigation, on lessons learned and benefits. Some of the investigative techniques discussed include process assessments, video inspections, hydrostatic and tracer tests, Helium 3 analysis for tritium age dating, deuterium and tritium in soil analysis. The investigative techniques have widespread applications to other nuclear generating stations. (author)

Development and applications of reactor noise analysis at Ontario Hydro`s CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Gloeckler, O [Ontario Hydro, Toronto, ON (Canada); Tulett, M V [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station

1996-12-31

In 1992 a program was initiated to establish reactor noise analysis as a practical tool for plant performance monitoring and system diagnostics in Ontario Hydro`s CANDU reactors. Since then, various CANDU-specific noise analysis applications have been developed and validated. The noise-based statistical techniques are being successfully applied as powerful troubleshooting and diagnostic tools to a wide variety of actual operational I and C problems. The dynamic characteristics of critical plant components, instrumentation and processes are monitored on a regular basis. Recent applications of noise analysis include (1) validating the dynamics of in-core flux detectors (ICFDS) and ion chambers, (2) estimating the prompt fraction ICFDs in noise measurements at full power and in power rundown tests, (3) identifying the cause of excessive signal fluctuations in certain flux detectors, (4) validating the dynamic coupling between liquid zone control signals, (5) detecting and monitoring mechanical vibrations of detector tubes induced by moderator flow, (6) estimating the dynamics and response time of RTD (Resistance Temperature Detector) temperature signals, (7) isolating the cause of RTD signal anomalies, (8) investigating the source of abnormal flow signal behaviour, (9) estimating the overall response time of flow and pressure signals, (10) detecting coolant boiling in fully instrumented fuel channels, (11) monitoring moderator circulation via temperature noise, and (12) predicting the performance of shut-off rods. Some of these applications are performed on an as-needed basis. The noise analysis program, in the Pickering-B station alone, has saved Ontario Hydro millions of dollars during its first three years. The results of the noise analysis program have been also reviewed by the regulator (Atomic Energy Control Board of Canada) with favorable results. The AECB have expressed interest in Ontario Hydro further exploiting the use of noise analysis technology. (author

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

Energy Technology Data Exchange (ETDEWEB)

Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

1995-09-01

This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

International Nuclear Information System (INIS)

Block, R.C.; Feiner, F.

1995-09-01

This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 2, Sessions 6-11

Energy Technology Data Exchange (ETDEWEB)

Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

1995-09-01

This document, Volume 2, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

Energy Technology Data Exchange (ETDEWEB)

Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

1995-09-01

This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

The Vale of Pickering in the Mesolithic: uncovering the early post-glacial landscape

Directory of Open Access Journals (Sweden)

Tim Schadla-Hall

2000-10-01

Full Text Available Since 1954, when Grahame Clark published the results of his excavations at Star Carr in northeast Yorkshire, the site has been recognized as a key to understanding early Mesolithic huntergatherer settlement and subsistence in northwest Europe. In 1976, archaeological and palaeoenvironmental research in the area was resumed - since 1986 under the auspices of the Vale of Pickering Research Trust - and it is now possible to set Star Carr and nearby Mesolithic sites in the wider context of the early postglacial landscape.

Assessment of ichthyoplankton entrainment at Pickering 'A' NGS using a pump/net in lake system

International Nuclear Information System (INIS)

McKinley, R.S.

1985-03-01

Annual entrainment at Pickering 'A' NGS was estimated for alewife as 13.6 X 10 6 larvae and 409 X 10 6 eggs. A substantial portion of eggs and larvae entering the intake were dead due to natural mortality (41%-81%) prior to entrainment. Viable eggs and larvae, immediately following entrainment showed mortalities of 54% and 44% respectively. The latent mortality of entrained eggs was 100% (48 h)

Remote tooling for inspection and repair in Pickering NGS-A calandria vault

International Nuclear Information System (INIS)

Hadji-Mirzai, M.; Tokarz, A.; Vandenberg, J.P.

1993-01-01

In recent years it has been necessary to develop capabilities for the inspection and repair of carbon steel components located within calandria vaults at Ontario Hydro's Pickering Nuclear Generating Station 'A'. Concerns about corrosion of piping and some of the structural components have made necessary the development of remote manipulators to inspect and repair carbon steel components within the vaults to ensure continued reliable operation of the units. Remote manipulators for this program have been designed to perform a number of inspection and repair tasks, and several versions have been developed to specialise in detailed inspection techniques and precision tooling module manipulation. (author)

NUKAB system use with the PICKER DYNA CAMERA II

International Nuclear Information System (INIS)

Collet, H.; Faurous, P.; Lehn, A.; Suquet, P.

Present-day data processing units connected to scintillation gamma cameras can make use of cabled programme or recorded programme systems. The NUKAB system calls on the latter technique. The central element of the data processing unit, connected to the PICKER DYNA CAMERA II output, consists of a DIGITAL PDP 8E computer with 12-bit technological words. The use of a 12-bit technological format restricts the possibilities of digitalisation, 64x64 images representing the practical limit. However the NUKAB system appears well suited to the processing of data from gamma cameras at present in service. The addition of output terminals of the tracing panel type should widen the possibilities of the system. It seems that the 64x64 format is not a handicap in view of the resolution power of the detectors [fr

Interfacial behaviour of sodium stearoyllactylate (SSL) as an oil-in-water pickering emulsion stabiliser.

Science.gov (United States)

Kurukji, D; Pichot, R; Spyropoulos, F; Norton, I T

2013-11-01

The ability of a food ingredient, sodium stearoyllactylate (SSL), to stabilise oil-in-water (O/W) emulsions against coalescence was investigated, and closely linked to its capacity to act as a Pickering stabiliser. Results showed that emulsion stability could be achieved with a relatively low SSL concentration (≥0.1 wt%), and cryogenic-scanning electron microscopy (cryo-SEM) visualisation of emulsion structure revealed the presence of colloidal SSL aggregates adsorbed at the oil-water interface. Surface properties of SSL could be modified by altering the size of these aggregates in water; a faster decrease in surface tension was observed when SSL dispersions were subjected to high pressure homogenisation (HPH). The rate of SSL adsorption at the sunflower oil-water interface also increased after HPH, and a higher interfacial tension (IFT) was observed with increasing SSL concentration. Differential scanning calorimetry (DSC) enabled a comparison of the thermal behaviour of SSL in aqueous dispersions with SSL-stabilised O/W emulsions. SSL melting enthalpy depended on emulsion interfacial area and the corresponding DSC data was used to determine the amount of SSL adsorbed at the oil-water interface. An idealised theoretical interfacial coverage calculation based on Pickering emulsion theory was in general agreement with the mass of SSL adsorbed as predicted by DSC. Copyright © 2013 The Authors. Published by Elsevier Inc. All rights reserved.

Nozzleless Fabrication of Oil-Core Biopolymeric Microcapsules by the Interfacial Gelation of Pickering Emulsion Templates.

Science.gov (United States)

Leong, Jun-Yee; Tey, Beng-Ti; Tan, Chin-Ping; Chan, Eng-Seng

2015-08-05

Ionotropic gelation has been an attractive method for the fabrication of biopolymeric oil-core microcapsules due to its safe and mild processing conditions. However, the mandatory use of a nozzle system to form the microcapsules restricts the process scalability and the production of small microcapsules (microcapsules through ionotropic gelation at the interface of an O/W Pickering emulsion. This approach involves the self-assembly of calcium carbonate (CaCO3) nanoparticles at the interface of O/W emulsion droplets followed by the addition of a polyanionic biopolymer into the aqueous phase. Subsequently, CaCO3 nanoparticles are dissolved by pH reduction, thus liberating Ca(2+) ions to cross-link the surrounding polyanionic biopolymer to form a shell that encapsulates the oil droplet. We demonstrate the versatility of this method by fabricating microcapsules from different types of polyanionic biopolymers (i.e., alginate, pectin, and gellan gum) and water-immiscible liquid cores (i.e., palm olein, cyclohexane, dichloromethane, and toluene). In addition, small microcapsules with a mean size smaller than 100 μm can be produced by selecting the appropriate conventional emulsification methods available to prepare the Pickering emulsion. The simplicity and versatility of this method allows biopolymeric microcapsules to be fabricated with ease by ionotropic gelation for numerous applications.

Ring thermal shield piping modification at Pickering Nuclear Generating Station 'A' Unit 1

International Nuclear Information System (INIS)

Brown, R.; Cobanoglu, M.M.

1995-01-01

Each of the four Pickering Nuclear Generating Station A (PNGSA) CANDU units was constructed with its reactor and dump tank surrounded by a concrete Calandria Vault (CV). The Ring Thermal Shield (RTS) system at PNGSA units is a water cooled structure with internal cooling channels with the purpose of attenuating excessive heat flux from the calandria shell to the end shield rings and adjoining concrete (Figure 1). In newer CANDU units the reactor calandria vessel is surrounded by a large water filled shield tank which eliminates the requirement for the RTS system. The RTS structures are situated in the space between the calandria and the vault walls. Each RTS is assembled from eight flat sided carbon steel segments, tilted towards the calandria and supported from the end shield rings. Cooling water to the RTS is supplied by carbon steel cooling pipes with a portion of the pipe run embedded in the vault walls. Flow through each RTS is divided into two independent circuits, having an inlet and an outlet cooling line. There are four locations of RTS inlet and outlet cooling lines. The inlet lines are located at the bottom and the outlet lines at the top of the RTS. The 'L' shaped section of RTS inlet and outlet cooling lines, from the RTS waterbox to the start of embedded portion at the concrete wall, had become defective due to corrosion induced by excessive Moisture levels in the calandria vaults. An on-line leak sealing capability was developed and placed in service in all four PNGSA units. However, a leak found during the 1994 Unit 1 outage was too large,to seal with the current capability, forcing Ontario Hydro (OH) to develop a method to replace the corroded pipes. The repair project was subject to some lofty performance targets. All tools had to be able to withstand dose rates of up to 3000 Rem/hour. These tools, along with procedures and personnel had to successfully repair the RTS system within 6 months otherwise a costly outage extension would result. This

Fabrication of CMC-g-PAM superporous polymer monoliths via eco-friendly Pickering-MIPEs for superior adsorption of methyl violet and methylene blue

Science.gov (United States)

Wang, Feng; Zhu, Yongfeng; Wang, Wenbo; Zong, Li; Lu, Taotao; Wang, Aiqin

2017-06-01

A series of superporous carboxymethylcellulose-graft-poly(acrylamide) (CMC-g-PAM) polymer monoliths presenting interconnected pore structure and excellent adsorption properties were prepared by one-step free-radical grafting polymerization reaction of CMC and acrylamide (AM) in the oil-in-water (O/W) Pickering-medium internal phase emulsions (Pickering-MIPEs) composed of non-toxic edible oil as a dispersion phase and natural Pal nanorods as stabilizers. The effects of Pal dosage, AM dosage, and co-surfactant Tween-20 (T-20) on the pore structures of the monoliths were studied. It was revealed that the well-defined pores were formed when the dosages of Pal and T-20 are 9-14% and 3%, respectively. The porous monolith can rapidly adsorb 1585 mg/g of methyl violet (MV) and 1625 mg/g of methylene blue (MB). After the monolith was regenerated by adsorption-desorption process for 5 times, the adsorption capacities still reached 92.1% (for MV) and 93.5% (for MB) of the initial maximum adsorption capacities. The adsorption process was fitted with Langmuir adsorption isotherm model and pseudo-second-order adsorption kinetic model very well, which indicate that mono-layer chemical adsorption mainly contribute to the high-capacity adsorption for dyes. The superporous polymer monolith prepared from eco-friendly Pickering-MIPEs shows good adsorption capacity and fast adsorption rate, which is potential adsorbent for the decontimination of dye-containing wastewater.

Reverse logistics network for municipal solid waste management: The inclusion of waste pickers as a Brazilian legal requirement.

Science.gov (United States)

Ferri, Giovane Lopes; Chaves, Gisele de Lorena Diniz; Ribeiro, Glaydston Mattos

2015-06-01

This study proposes a reverse logistics network involved in the management of municipal solid waste (MSW) to solve the challenge of economically managing these wastes considering the recent legal requirements of the Brazilian Waste Management Policy. The feasibility of the allocation of MSW material recovery facilities (MRF) as intermediate points between the generators of these wastes and the options for reuse and disposal was evaluated, as well as the participation of associations and cooperatives of waste pickers. This network was mathematically modelled and validated through a scenario analysis of the municipality of São Mateus, which makes the location model more complete and applicable in practice. The mathematical model allows the determination of the number of facilities required for the reverse logistics network, their location, capacities, and product flows between these facilities. The fixed costs of installation and operation of the proposed MRF were balanced with the reduction of transport costs, allowing the inclusion of waste pickers to the reverse logistics network. The main contribution of this study lies in the proposition of a reverse logistics network for MSW simultaneously involving legal, environmental, economic and social criteria, which is a very complex goal. This study can guide practices in other countries that have realities similar to those in Brazil of accelerated urbanisation without adequate planning for solid waste management, added to the strong presence of waste pickers that, through the characteristic of social vulnerability, must be included in the system. In addition to the theoretical contribution to the reverse logistics network problem, this study aids in decision-making for public managers who have limited technical and administrative capacities for the management of solid wastes. Copyright © 2015 Elsevier Ltd. All rights reserved.

Eddy current magnetic bias x-probe qualification and inspection of steam generator Monel 400 tubing in Pickering Nuclear Generating Station

International Nuclear Information System (INIS)

Lepine, B.A.; Van Langen, J.; Obrutsky, L.

2006-01-01

This paper presents an overview of the x-probe MB 350 eddy current inspection array probe, for detection of open OD axial crack-like flaws in Monel 400 tubes at Pickering Nuclear Generating Station. This report contains a selection of inspection results from the field inspections performed with this probe during the 2003 and 2004 period at Pickering Nuclear Generating Station A and B. During the 2003 in-service eddy current inspection results of Pickering Nuclear Generating Station A (PNGS-A) Unit 2, a 13 mm (0.5 inch) long axial indication was detected by the CTR1 bobbin and CTR2-C4 array probes in Tube R25-C52 of Steam Generator (SG) 11 in the hot leg sludge pile region. An experimental magnetic bias X-probe, specially designed by Zetec for inspection of Monel 400 tubing, was deployed and the indication was characterized as a potential out diameter (OD) axially oriented crack. Post-inspection tube pulling and destructive examination confirmed the presence of an Environmentally Assisted Crack (EAC), approximately 80% deep and 13mm long. Due to the significance of this discovery, Ontario Power Generation (OPG) requested AECL to initiate a program for qualification of the X-probe MB 350 for the detection of OD axial cracks in medium to high magnetic permeability μ r Monel 400 PNGS-A and B steam generator tubing at different locations. The X-probe MB 350 subsequently has been deployed as a primary inspection probe for crack detection for PNGS steam generators. (author)

Modulation of Cyclodextrin Particle Amphiphilic Properties to Stabilize Pickering Emulsion.

Science.gov (United States)

Xi, Yongkang; Luo, Zhigang; Lu, Xuanxuan; Peng, Xichun

2018-01-10

Cyclodextrins have been proven to form complexes with linear oil molecules and stabilize emulsions. Amphiphilic properties of cyclodextrin particles were modulated through esterification reaction between β-cyclodextrin (β-CD) and octadecenyl succinic anhydride (ODSA) under alkaline conditions. ODS-β-CD particles with degree of substitution (DS) of 0.003, 0.011, and 0.019 were obtained. The introduced hydrophobic long chain that was linked within β-CD cavity led to the change of ODS-β-CD in terms of morphological structure, surface charge density, size, and contact angle, upon which the properties and stability of the emulsions stabilized by ODS-β-CD were highly dependent. The average diameter of ODS-β-CD particles ranged from 449 to 1484 nm. With the DS increased from 0.003 to 0.019, the contact angle and absolute zeta potential value of these ODS-β-CD particles improved from 25.7° to 47.3° and 48.1 to 62.8 mV, respectively. The cage structure of β-CD crystals was transformed to channel structure, then further to amorphous structure after introduction of the octadecenyl succinylation chain. ODS-β-CD particles exhibited higher emulsifying ability compared to β-CD. The resulting Pickering emulsions formed by ODS-β-CD particles were more stable during storage. This study investigates the ability of these ODS-β-CD particles to stabilize oil-in-water emulsions with respect to their amphiphilic character and structural properties.

Electrospun composite matrices of poly(ε-caprolactone)-montmorillonite made using tenside free Pickering emulsions

International Nuclear Information System (INIS)

Samanta, Archana; Takkar, Sonam; Kulshreshtha, Ritu; Nandan, Bhanu; Srivastava, Rajiv K.

2016-01-01

The production of composite electrospun matrices of poly(ε-caprolactone) (PCL) using an emulsifier-free emulsion, made with minimal organic solvent, as precursor is reported. Pickering emulsions of PCL were prepared using modified montmorillonite (MMT) clay as the stabilizer. Hydrophobic tallow group of the modified MMT clay resulted in analogous interaction of clay with oil and aqueous phase and its adsorption at the interface to provide stability to the resultant emulsion. Composite fibrous matrices of PCL and MMT were produced using electrospinning under controlled conditions. The fiber fineness was found to alter with PCL concentration and volume fraction of the aqueous and oil phases. A higher tensile strength and modulus was obtained with inclusion of MMT in PCL electrospun matrix in comparison to a matrix made using neat PCL. The presence of clay in the fibrous matrix did not change the cell proliferation efficiency in comparison to neat PCL matrix. Composite fibrous matrices of PCL/MMT bearing enhanced tensile properties may find applications in areas other than tissue engineering for example food packaging and filtration. - Highlights: • Tenside free, clay stabilized Pickering emulsion of PCL is made with minimal organic solvent. • Organic–inorganic composite fibrous matrices were produced via emulsion electrospinning. • Fiber fineness was efficiently controlled by variation in emulsion formulation. • Fibrous matrices of high tensile strength and modulus were obtained in comparison to neat PCL matrix. • PCL/clay matrices showed effective cell proliferation as a neat PCL matrix.

Electrospun composite matrices of poly(ε-caprolactone)-montmorillonite made using tenside free Pickering emulsions

Energy Technology Data Exchange (ETDEWEB)

Samanta, Archana [Department of Textile Technology, Indian Institute of Technology Delhi, Hauz Khas, New Delhi 110016 (India); Takkar, Sonam; Kulshreshtha, Ritu [Department of Biochemical Engineering and Biotechnology, Indian Institute of Technology Delhi, Hauz Khas, New Delhi 110016 (India); Nandan, Bhanu [Department of Textile Technology, Indian Institute of Technology Delhi, Hauz Khas, New Delhi 110016 (India); Srivastava, Rajiv K., E-mail: rajiv@textile.iitd.ac.in [Department of Textile Technology, Indian Institute of Technology Delhi, Hauz Khas, New Delhi 110016 (India)

2016-12-01

The production of composite electrospun matrices of poly(ε-caprolactone) (PCL) using an emulsifier-free emulsion, made with minimal organic solvent, as precursor is reported. Pickering emulsions of PCL were prepared using modified montmorillonite (MMT) clay as the stabilizer. Hydrophobic tallow group of the modified MMT clay resulted in analogous interaction of clay with oil and aqueous phase and its adsorption at the interface to provide stability to the resultant emulsion. Composite fibrous matrices of PCL and MMT were produced using electrospinning under controlled conditions. The fiber fineness was found to alter with PCL concentration and volume fraction of the aqueous and oil phases. A higher tensile strength and modulus was obtained with inclusion of MMT in PCL electrospun matrix in comparison to a matrix made using neat PCL. The presence of clay in the fibrous matrix did not change the cell proliferation efficiency in comparison to neat PCL matrix. Composite fibrous matrices of PCL/MMT bearing enhanced tensile properties may find applications in areas other than tissue engineering for example food packaging and filtration. - Highlights: • Tenside free, clay stabilized Pickering emulsion of PCL is made with minimal organic solvent. • Organic–inorganic composite fibrous matrices were produced via emulsion electrospinning. • Fiber fineness was efficiently controlled by variation in emulsion formulation. • Fibrous matrices of high tensile strength and modulus were obtained in comparison to neat PCL matrix. • PCL/clay matrices showed effective cell proliferation as a neat PCL matrix.

Fragger: a protein fragment picker for structural queries.

Science.gov (United States)

Berenger, Francois; Simoncini, David; Voet, Arnout; Shrestha, Rojan; Zhang, Kam Y J

2017-01-01

Protein modeling and design activities often require querying the Protein Data Bank (PDB) with a structural fragment, possibly containing gaps. For some applications, it is preferable to work on a specific subset of the PDB or with unpublished structures. These requirements, along with specific user needs, motivated the creation of a new software to manage and query 3D protein fragments. Fragger is a protein fragment picker that allows protein fragment databases to be created and queried. All fragment lengths are supported and any set of PDB files can be used to create a database. Fragger can efficiently search a fragment database with a query fragment and a distance threshold. Matching fragments are ranked by distance to the query. The query fragment can have structural gaps and the allowed amino acid sequences matching a query can be constrained via a regular expression of one-letter amino acid codes. Fragger also incorporates a tool to compute the backbone RMSD of one versus many fragments in high throughput. Fragger should be useful for protein design, loop grafting and related structural bioinformatics tasks.

Fabrication of CMC-g-PAM Superporous Polymer Monoliths via Eco-Friendly Pickering-MIPEs for Superior Adsorption of Methyl Violet and Methylene Blue.

Science.gov (United States)

Wang, Feng; Zhu, Yongfeng; Wang, Wenbo; Zong, Li; Lu, Taotao; Wang, Aiqin

2017-01-01

A series of superporous carboxymethylcellulose- graft -poly(acrylamide)/palygorskite (CMC- g -PAM/Pal) polymer monoliths presenting interconnected pore structure and excellent adsorption properties were prepared by one-step free-radical grafting polymerization reaction of CMC and acrylamide (AM) in the oil-in-water (O/W) Pickering-medium internal phase emulsions (Pickering-MIPEs) composed of non-toxic edible oil as a dispersion phase and natural Pal nanorods as stabilizers. The effects of Pal dosage, AM dosage, and co-surfactant Tween-20 (T-20) on the pore structures of the monoliths were studied. It was revealed that the well-defined pores were formed when the dosages of Pal and T-20 are 9-14 and 3%, respectively. The porous monolith can rapidly adsorb 1,585 mg/g of methyl violet (MV) and 1,625 mg/g of methylene blue (MB). After the monolith was regenerated by adsorption-desorption process for five times, the adsorption capacities still reached 92.1% (for MV) and 93.5% (for MB) of the initial maximum adsorption capacities. The adsorption process was fitted with Langmuir adsorption isotherm model and pseudo-second-order adsorption kinetic model very well, which indicate that mono-layer chemical adsorption mainly contribute to the high-capacity adsorption for dyes. The superporous polymer monolith prepared from eco-friendly Pickering-MIPEs shows good adsorption capacity and fast adsorption rate, which is potential adsorbent for the decontamination of dye-containing wastewater.

Fabrication of CMC-g-PAM Superporous Polymer Monoliths via Eco-Friendly Pickering-MIPEs for Superior Adsorption of Methyl Violet and Methylene Blue

Directory of Open Access Journals (Sweden)

Feng Wang

2017-06-01

Full Text Available A series of superporous carboxymethylcellulose-graft-poly(acrylamide/palygorskite (CMC-g-PAM/Pal polymer monoliths presenting interconnected pore structure and excellent adsorption properties were prepared by one-step free-radical grafting polymerization reaction of CMC and acrylamide (AM in the oil-in-water (O/W Pickering-medium internal phase emulsions (Pickering-MIPEs composed of non-toxic edible oil as a dispersion phase and natural Pal nanorods as stabilizers. The effects of Pal dosage, AM dosage, and co-surfactant Tween-20 (T-20 on the pore structures of the monoliths were studied. It was revealed that the well-defined pores were formed when the dosages of Pal and T-20 are 9–14 and 3%, respectively. The porous monolith can rapidly adsorb 1,585 mg/g of methyl violet (MV and 1,625 mg/g of methylene blue (MB. After the monolith was regenerated by adsorption-desorption process for five times, the adsorption capacities still reached 92.1% (for MV and 93.5% (for MB of the initial maximum adsorption capacities. The adsorption process was fitted with Langmuir adsorption isotherm model and pseudo-second-order adsorption kinetic model very well, which indicate that mono-layer chemical adsorption mainly contribute to the high-capacity adsorption for dyes. The superporous polymer monolith prepared from eco-friendly Pickering-MIPEs shows good adsorption capacity and fast adsorption rate, which is potential adsorbent for the decontamination of dye-containing wastewater.

Status of IVO-FR2-Vg7 experiment for irradiation of fast reactor fuel rods

International Nuclear Information System (INIS)

Elbel, H.; Kummerer, K.; Bojarsky, K.; Lopez Jimenez, J.; Otero de la Gandara, J.L.

1979-01-01

Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid) concerning a Joint Irradiation Program of Fast Reactor Fuel Rods. The design of fuel rods in general is defined, and, in particular of those with a density 94% DT and diameter 7.6 mm up to a burn-up of 7% FIMA, to be irradiated in the FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used in this irradiation, other possibilities of irradiation in the reactor will also be described. (auth.)

Exploring the role of picker personality in predicting picking performance with pick by voice, pick to light and RF-terminal picking

NARCIS (Netherlands)

de Vries, J.; de Koster, R.; Stam, D.

2016-01-01

Order pickers and individual differences between them could have a substantial impact on picking performance, but are largely ignored in studies on order picking. This paper explores the role of individual differences in picking performance with various picking tools (pick by voice, RF-terminal

Preparation of stable Pickering emulsions with short, medium and long chain fats and starch nanocrystals and their in vitro digestion properties

Science.gov (United States)

Pickering emulsions are receiving more attention as delivery systems in food and pharmaceuticals because they can be formulated with nontoxic food ingredients to form stable emulsions. In this study, 40-100 nm starch nanocrystals (SNCs) prepared from acid hydrolysis of waxy maize starches were used ...

Maintenance of ageing CANDU reactors. A regulatory perspective

International Nuclear Information System (INIS)

Dunstan, T.

1996-01-01

The subject of this paper is, 'requirements for maintenance of ageing reactors from the perspective of a regulator', with a focus on the particular theme of; 'continuing safety assurance'. A major role of maintenance is to ensure the continuing reliability and effectiveness of safety related systems and equipment. Continuing safety assurance is an issue the Atomic Energy Control Board has been wrestling with for some time. From my perspective, much remains to be done before the AECB can be confident that Canadian nuclear plants have the necessary programs in place to achieve continuing safety assurance. To introduce the topic, it would be appropriate to say a few words about the AECB's position with respect to the situation at the Pickering NGS. Why did we blow the whistle last August and, what are we doing about it? (author)

Tritium in groundwater investigation at the Pickering Nuclear Generating Station

International Nuclear Information System (INIS)

DeWilde, J.; Yu, L.; Wootton, R.; Belanger, D.; Hansen, K.; McGurk, E.; Teare, A.

2001-01-01

Ontario Power Generation Inc. (OPG) investigated tritium in groundwater at the Pickering Nuclear Generating Station (PNGS). The objectives of the study were to evaluate and define the extent of radionuclides, primarily tritium, in groundwater, investigate the causes or sources of contamination, determine impacts on the natural environment, and provide recommendations to prevent future discharges. This paper provides an overview of the investigations conducted in 1999 and 2000 to identity the extent of the tritium beneath the site and the potential sources of tritium released to the groundwater. The investigation and findings are summarized with a focus on unique aspects of the investigation, on lessons learned and benefits. Some of the investigative techniques discussed include process assessments, video inspections, hydrostatic and tracer tests, Helium 3 analysis for tritium age dating, deuterium and tritium in soil analysis. The investigative techniques have widespread applications to other nuclear generating stations. (author)

A progress review of Ontario Hydro's nuclear generation and heavy water production programs

International Nuclear Information System (INIS)

Kee, F.J.; Woodhead, L.W.

Performance and economics of CANDU reactors in service are described. Progress of commissioning, construction and planning of reactors at Pickering, Bruce, and Darlington is outlined. Heavy water production is reviewed. (E.C.B.)

Enhanced decolourisation of Acid Orange 7 in a continuous UASB reactor with quinones as redox mediators.

NARCIS (Netherlands)

Cervantes, F.J.; Zee, van der F.P.; Lettinga, G.; Field, J.A.

2001-01-01

The reductive biotransformation of acid orange 7 (AO7) was explored in a lab-scale upflow anaerobic sludge blanket (UASB) reactor at low hydraulic residence times (HRT). A colour removal of 85% was achieved when the reactor was operated at a HRT of 6 hours, but decreased up to 70% when the HRT was

Quebec Gentilly 2 nuclear power station

International Nuclear Information System (INIS)

Labbe, J.A.

Modifications and commissioning of the Gentilly reactor are described. The Gentilly reactor is owned by AECL, not Quebec Hydro, and has served as a prototype reactor. The Gentilly-2 reactor is a 'packaged' 600 MWe PHW reactor similar to Pickering-1, etc. Interesting aspects of construction and purchasing of equipment are described. (E.C.B.)

Seismic response of the Pickering pressure relief duct to the 1985 Nahanni earthquake

International Nuclear Information System (INIS)

Ghobarah, A.

1995-05-01

The objective of this study is to examine the structural response of the Pickering pressure relief duct when subjected to the ground motion records of the 1985 Nahanni earthquake (December 23, 05:16 GMT, Site 1 - Iverson, N.W.T.). It also includes an estimate of the possible impact on the nuclear safety function of the duct. The structural models developed in an earlier study were used in this analysis. The response to the earthquake ground motion was determined on the basis of the estimated capacities of various components of the duct. The ability of the structure to fulfill its nuclear safety function is discussed. (author). 6 refs., 1 tab., 17 figs

Molecularly imprinted polymer microspheres prepared by Pickering emulsion polymerization for selective solid-phase extraction of eight bisphenols from human urine samples

International Nuclear Information System (INIS)

Yang, Jiajia; Li, Yun; Wang, Jincheng; Sun, Xiaoli; Cao, Rong; Sun, Hao; Huang, Chaonan; Chen, Jiping

2015-01-01

Highlights: • BPA imprinted polymer microspheres were prepared by Pickering emulsion polymerization. • Regular spherical shape and narrow diameter distribution. • Good specific adsorption capacity for BPA. • Good class-selectivity and clean-up efficiency for bisphenols in human urine under SPE mode. • Good recoveries and sensitivity for bisphenols using the MIPMS-SPE coupled with HPLC-DAD method. - Abstract: The bisphenol A (BPA) imprinted polymer microspheres were prepared by simple Pickering emulsion polymerization. Compared to traditional bulk polymerization, both high yields of polymer and good control of particle sizes were achieved. The characterization results of scanning electron microscopy and nitrogen adsorption–desorption measurements showed that the obtained molecularly imprinted polymer microsphere (MIPMS) particles possessed regular spherical shape, narrow diameter distribution (30–60 μm), a specific surface area (S BET ) of 281.26 m 2 g −1 and a total pore volume (V t ) of 0.459 cm 3 g −1 . Good specific adsorption capacity for BPA was obtained in the sorption experiment and good class selectivity for BPA and its seven structural analogs (bisphenol F, bisphenol B, bisphenol E, bisphenol AF, bisphenol S, bisphenol AP and bisphenol Z) was demonstrated by the chromatographic evaluation experiment. The MIPMS as solid-phase extraction (SPE) packing material was then evaluated for extraction and clean-up of these bisphenols (BPs) from human urine samples. An accurate and sensitive analytical method based on the MIPMS-SPE coupled with HPLC-DAD has been successfully established for simultaneous determination of eight BPs from human urine samples with detection limits of 1.2–2.2 ng mL −1 . The recoveries of BPs for urine samples at two spiking levels (100 and 500 ng mL −1 for each BP) were in the range of 81.3–106.7% with RSD values below 8.3%

Pickering emulsion stabilized by cashew gum- poly-l-lactide copolymer nanoparticles: Synthesis, characterization and amphotericin B encapsulation.

Science.gov (United States)

Richter, A R; Feitosa, J P A; Paula, H C B; Goycoolea, F M; de Paula, R C M

2018-04-01

In this work, we provide proof-of-concept of formation, physical characteristics and potential use as a drug delivery formulation of Pickering emulsions (PE) obtained by a novel method that combines nanoprecipitation with subsequent spontaneous emulsification process. To this end, pre-formed ultra-small (d.∼10 nm) nanoprecipitated nanoparticles of hydrophobic derivatives of cashew tree gum grafted with polylactide (CGPLAP), were conceived to stabilize Pickering emulsions obtained by spontaneous emulsification. These were also loaded with Amphotericin B (AmB), a drug of low oral bioavailability used in the therapy of neglected diseases such as leishmaniasis. The graft reaction was performed in two CG/PLA molar ratio conditions (1:1 and 1:10). Emulsions were prepared by adding the organic phase (Miglyol 812 ® ) in the aqueous phase (nanoprecipitated CGPLAP), resulting the immediate emulsion formation. The isolation by centrifugation does not destabilize or separate the nanoparticles from oil droplets of the PE emulsion. Emulsions with CGPLAP 1:1 presented unimodal distributions at different CGPLA concentration, lower values in size and PDI and the best stability over time. The AmB was incorporated in the emulsions with a process efficiency of 21-47%, as determined by UV-vis. AmB in CGPLAP emulsions is in less aggregated state than observed in commercial AmB formulation. Copyright © 2018 Elsevier B.V. All rights reserved.

Radiological pathways analysis for spent solvents from the boiler chemical cleaning at the Pickering Nuclear Site

International Nuclear Information System (INIS)

Garisto, N.C.; Eslami, Z.; Hodgins, S.; Beaman, T.; Von Svoboda, S.; Marczak, J.

2006-01-01

Spent solvents are generated as a result of Boiler Chemical Cleanings (BCC) at CANDU reactor sites. These solutions contain small amount of radioactivity from a number of different sources including: Cut tubes - short sections of boiler tubes are infrequently removed from the boilers for a detailed characterization. These tubes are typically only plugged at the tubesheet allowing the primary side deposits to be exposed to BCC solvents. Tube leaks - primary to secondary side leaks also occur infrequently as a result of tube degradation. Radioactivity from the leaking fluid can consequently be deposited in the sludge on the secondary side of the tubes. Diffusion of tritium - during normal operation of the reactor units, tritium slowly diffuses from the heavy water in the primary heat-transfer system to the light-water coolant on the secondary side. Some of this tritium is retained in the secondary side deposits. The Pickering Nuclear Generating Station (PNGS) would like the flexibility to have several options for handling the spent solvent waste and associated rinse water from BCC. To this end, a radiological pathways analysis was undertaken to determine dose consequences associated with each option. Sample results from this study are included in this paper. The pathways analysis is used in this study to calculate dose to hypothetical receptors including individuals such as truck drivers, incinerator workers, residue (ash) handlers, residents who live near the landfill, inadvertent intruders into the landfill after closure and residents who live near the outfall. This dose is compared to a de minimis dose. A de minimis dose or dose rate represents a level of risk, which is generally accepted as being of no significance. Shipments of spent solvents and rinse water with corresponding doses below de minimis can be sent to conventional (i.e., non-radioactive) landfills for incineration and disposal as the radioactive dose associated with them is much less than natural

Cobalt-60 production in CANDU power reactors

International Nuclear Information System (INIS)

Malkoske, G.R.; Norton, J.L.; Slack, J.

2002-01-01

MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments, but are primarily used to sterilize single-use medical products including; surgical kits, gloves, gowns, drapes, and cotton swabs. Other applications include sanitization of cosmetics, microbial reduction of pharmaceutical raw materials, and food irradiation. The technology for producing the cobalt-60 isotope was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) almost 55 years ago using research reactors at the AECL Chalk River Laboratories in Ontario, Canada. The first cobalt-60 source produced for medical applications was manufactured by MDS Nordion and used in cancer therapy. The benefits of cobalt-60 as applied to medical product manufacturing, were quickly realized and the demand for this radioisotope quickly grew. The same technology for producing cobalt-60 in research reactors was then designed and packaged such that it could be conveniently transferred to a utility/power reactor. In the early 1970's, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production for industrial irradiation applications was initiated in the four Pickering A CANDU reactors. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology for producing cobalt-60 in additional CANDU reactors. CANDU is unique among the power reactors of the world, being heavy water moderated and fuelled with natural uranium. They are also designed and supplied with stainless steel adjusters, the primary function of which is to shape the neutron flux to optimize reactor power and fuel bum-up, and to provide excess reactivity needed to overcome xenon-135 poisoning following a reduction of power. The reactor is designed to develop full power output with all of the adjuster

The Effect of Storage and Routing Policies on Picker Blocking in a Real-life Narrow-aisle Warehouse

OpenAIRE

Van Gils, Teun; Caris, An; Ramaekers, Katrien

2017-01-01

Upcoming e-commerce markets force warehouses to handle a large number of orders within short time windows. Narrow-aisle order picking systems allow to store a large number of products in small areas. In manual order picking systems, narrow aisles can result in substantial waiting time compared to wide-aisle systems. The objective of this study is to analyse the joint effect of the two main operational order picking planning problems, storage location assignment and order picker routing, on or...

Characteristics of outage radiation fields around various reactor components

International Nuclear Information System (INIS)

Verzilov, Y.; Husain, A.; Corbin, G.

2008-01-01

Full text: Activity monitoring surveys, consisting of gamma spectroscopy and dose rate measurements, of various CANDU station components such as the reactor face, feeder cabinet, steam generators and moderator heat exchangers are often performed during shutdown in order to trend the transport of activity around the primary heat transport and moderator systems. Recently, the increased dose expenditure for work such as feeder inspection and replacement in the reactor vault has also spurred interest in improved characterization of the reactor face fields to facilitate better ALARA decision making and hence a reduction in future dose expenditures. At present, planning for reactor face work is hampered by insufficient understanding of the relative contribution of the various components to the overall dose. In addition to the increased dose expenditure for work at the reactor face, maintenance work associated with horizontal flux detectors and liquid injection systems has also resulted in elevated dose expenditures. For instance at Darlington, radiation fields in the vicinity of horizontal flux detectors (HFD) and Liquid Injection Shutdown System (LISS) nozzle bellows are trending upwards with present contact fields being in the range 16-70 rem/h and working distance fields being in the range 100-500 mrem/h. This paper presents findings based on work currently being funded by the CANDU Owners Group. Measurements were performed at Ontario Power Generation's Pickering and Darlington nuclear stations. Specifically, the following are addressed: Characteristics of Reactor Vault Fields; Characteristics of Steam Generator Fields; Characteristics of Moderator Heat Exchanger Fields. Measurements in the reactor vault were performed at the reactor face, along the length of end fittings, along the length of feeders, at the bleed condenser and at the HFD and LISS nozzle bellows. Steam generator fields were characterized at various elevations above the tube sheet, with and without the

Summary report of the 7th reduced-moderation water reactor workshop

International Nuclear Information System (INIS)

Akie, Hiroshi; Nabeshima, Kunihiko; Uchikawa, Sadao

2005-08-01

As a research on the future innovative water reactor, the development of Reduced-Moderation Water Reactors (RMWRs) has been performed in Japan Atomic Energy Research Institute (JAERI). The workshop on RMWRs is aiming at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors, and has been held every year since 1998. The 7th workshop was held on March 5, 2004 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The program of the workshop was composed of 5 lectures and an overall discussion time. The workshop started with the lecture by JAERI on the status and future program of PMWR research and development, followed by the two presentations by JAERI and Japan Nuclear Cycle Development Institute, respectively, on the investigation and evaluation of water cooled reactor in Feasibility Study Program on Commercialized Fast Reactor Systems. The lectures were also made on the Japan's nuclear fuel cycle and scenarios for RMWRs deployment by JAERI, and on the next generation reactor development activity by Hitachi, Ltd. The main subjects of the overall discussion time were Na cooled fast reactor, deployment effects of RMWRs and the future plan of the RMWR research and development. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as of the discussion time. In addition in the Appendices, there are included presentation handouts of each lecture, program of the workshop and the participants list. (author)

Interconnectivity of macroporous molecularly imprinted polymers fabricated by hydroxyapatite-stabilized Pickering high internal phase emulsions-hydrogels for the selective recognition of protein.

Science.gov (United States)

Sun, Yanhua; Li, Yuqing; Xu, Jiangfeng; Huang, Ling; Qiu, Tianyun; Zhong, Shian

2017-07-01

Hydroxyapatite hybridized molecularly imprinted polydopamine polymers with selective recognition of bovine hemoglobin (BHb) were successfully prepared via Pickering oil-in-water high internal phase emulsions-hydrogels and molecularly imprinting technique. The emulsions were stabilized by hydroxyapatite of which the wettability was modified by 3-methacryloxypropyltrimethoxysilane. The materials were characterized by SEM, IR and TGA. The results showed that the BHb imprinted polymers based on Pickering hydrogels (Hydro-MIPs) possess macropores ranging from 20μm to 50μm, and their large numbers of amino groups and hydroxyl groups result in a favorable adsorption capacity for BHb. The maximum adsorption capacity of Hydro-MIPs for BHb was 438mg/g, 3.27 times more than that of the non-imprinted polymers (Hydro-NIPs). The results indicated that Hydro-MIPs possessing well-defined hierarchical porous structures exhibited outstanding recognition behavior towards the target protein molecules. This work provided a promising alternative method for the fabrication of polymer materials with tunable and interconnected pores structures for the separation and purification of protein in vitro. Copyright © 2017. Published by Elsevier B.V.

A comparison of clinical vs subclinical skin pickers in Israel.

Science.gov (United States)

Keuthen, Nancy J; Curley, Erin E; Tung, Esther S; Ittah, Karen; Qasem, Atheer; Murad, Sari; Odlaug, Brian L; Leibovici, Vera

2016-05-01

Skin-picking disorder (SPD) was recognized as its own entity for the first time in DSM-5. The existing SPD literature is limited and, to date, no study has examined the differences between clinical and sub- clinical SPD. Identifying differences between these 2 groups may improve diagnostic accuracy, treatment, and prevention efforts. Israeli adults (N = 4,325) from 2 previous studies were examined for the presence of clinical and subclinical SPD. Individuals with clinical SPD (n = 150) vs subclinical SPD (n = 219) were compared on skin-picking characteristics, psychological phenomena, and clinical correlates. There were many similarities between clinical and subclinical skin pickers. Individuals with clinical SPD, however, had more severe skin picking, greater associated functional impairment, greater perceived stress, and greater depressive and obsessive-compulsive symptoms, and were also more likely to have a first-degree relative with SPD. This study suggests that although there are some similarities between clinical and subclinical SPD, there also are distinct differences in the clinical presentation. Understanding these differences may be an important factor in treatment and prevention planning.

Laboratory testing and assessment of the Pickering PRD supporting frame

International Nuclear Information System (INIS)

Ghobarah, A.

1995-05-01

The objective of this study was to design and test reinforced concrete beam-column subassemblages representing the beam, column and joint of the Centre Pier (CP) support of the Pressure Relief Duct (PRD) at the Pickering A Nuclear Generating Station. The testing program was expected to establish the failure mode of the subassemblage and to compare the performance of the existing CP with a specimen detailed in accordance with current code provisions. A one-third scale specimen of the beam-column subassemblage was designed and tested to failure when subjected to simulated seismic loads. A second specimen was constructed with shear reinforcement that was detailed according to the provisions of the CAN3-N287.3-M82 code. The second specimen was tested in the same manner as the first specimen. From the experimental data on the behaviour and mode of failure of the specimens, analytical evaluations were conducted to determine the inelastic nonlinear behaviour of the CP structural system when subjected to various levels of ground motion. (author). 11 refs., 3 tabs., 40 figs

Research reactors - an overview

International Nuclear Information System (INIS)

West, C.D.

1997-01-01

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

An approach of raising the low power reactor trip block (P-7) in Maanshan Power Plant

International Nuclear Information System (INIS)

Wang, L.C.

1984-01-01

The technical specification for the Maanshan Nuclear Power Station (FSAR Table 16.2.2-3) requires that with an increasing reactor power level above the setpoint of low power reactor trip block (P-7), a turbine trip shall initiate a reactor trip. This anticipatory reactor trip on turbine trip prevents the pressurizer PORV from openning during turbine trip event. In order to reduce unnecessary reactor trip due to turbine trip on low reactor power level during Maanshan start-up stage, Taiwan Power Company performed a transient analysis for turbine trip event by using RETRAN code. The highest reactor power level at which a turbine trip will not open the pressurizer PORV is searched. The results demonstrated that this power level can be increased from the original value-10% of the rated thermal power-to about 48% of the rated thermal power

CANDU fuel - fifteen years of power reactor experience

International Nuclear Information System (INIS)

Fanjoy, G.R.; Bain, A.S.

1977-01-01

CANDU (Canada Deuterium Uranium) fuel has operated in power reactors since 1962. Analyses of performance statistics, supplemented by examinations of fuel from power reactors and experimental loops have yielded: (a) A thorough understanding of the fundamental behaviour of CANDU fuel. (b) Data showing that the predicted high utilization of uranium has been achieved. Actual fuelling costs in 1976 at the Pickering Generating Station are 1.2 m$/kWh (1976 Canadian dollars) with the simple oncethrough natural-UO 2 fuel cycle. (c) Criteria for operation, which have led to the current very low defect rate of 0.03% of all assemblies and to ''CANLUB'' fuel, which has a graphite interlayer between the fuel and sheath to reduce defects on power increases. (d) Proof that the short length (500 mm), collapsible cladding features of the CANDU bundle are successful and that the fuel can operate at high-power output (current peak outer-element linear power is 58 +- 15% kW/m). Involvement by the utility in all stages of fuel development has resulted in efficient application of this fundamental knowledge to ensure proper fuel specifications, procurement, scheduling into the reactor and feedback to developers, designers and manufacturers. As of mid-1976 over 3 x 10 6 individual elements have been built in a well-estabilished commercially competitive fuel fabrication industry and over 2 x 10 6 elements have been irradiated. Only six defects have been attributed to faulty materials or fabrication, and the use of high-density UO 2 with low-moisture content precluded defects from hydrogen contamination and densification. Development work on UO 2 and other fuel cycles (plutonium and thorium) is continuing, and, because CANDU reactors use on-power fuelling, bundles can be inserted into power reactors for testing. Thus new fuel designs can be quickly adopted to ensure that the CANDU system continues to provide low-cost energy with high reliability

Molecularly imprinted polymer microspheres prepared by Pickering emulsion polymerization for selective solid-phase extraction of eight bisphenols from human urine samples

Energy Technology Data Exchange (ETDEWEB)

Yang, Jiajia [Key Laboratory of Separation Sciences for Analytical Chemistry, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, 457 Zhongshan Road, Dalian 116023 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Li, Yun; Wang, Jincheng [Key Laboratory of Separation Sciences for Analytical Chemistry, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, 457 Zhongshan Road, Dalian 116023 (China); Sun, Xiaoli; Cao, Rong [Key Laboratory of Separation Sciences for Analytical Chemistry, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, 457 Zhongshan Road, Dalian 116023 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Sun, Hao [Key Laboratory of Separation Sciences for Analytical Chemistry, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, 457 Zhongshan Road, Dalian 116023 (China); Department of Chemistry, Liaoning University, Shenyang 110000 (China); Huang, Chaonan [Key Laboratory of Separation Sciences for Analytical Chemistry, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, 457 Zhongshan Road, Dalian 116023 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Chen, Jiping, E-mail: chenjp@dicp.ac.cn [Key Laboratory of Separation Sciences for Analytical Chemistry, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, 457 Zhongshan Road, Dalian 116023 (China)

2015-05-04

Highlights: • BPA imprinted polymer microspheres were prepared by Pickering emulsion polymerization. • Regular spherical shape and narrow diameter distribution. • Good specific adsorption capacity for BPA. • Good class-selectivity and clean-up efficiency for bisphenols in human urine under SPE mode. • Good recoveries and sensitivity for bisphenols using the MIPMS-SPE coupled with HPLC-DAD method. - Abstract: The bisphenol A (BPA) imprinted polymer microspheres were prepared by simple Pickering emulsion polymerization. Compared to traditional bulk polymerization, both high yields of polymer and good control of particle sizes were achieved. The characterization results of scanning electron microscopy and nitrogen adsorption–desorption measurements showed that the obtained molecularly imprinted polymer microsphere (MIPMS) particles possessed regular spherical shape, narrow diameter distribution (30–60 μm), a specific surface area (S{sub BET}) of 281.26 m{sup 2} g{sup −1} and a total pore volume (V{sub t}) of 0.459 cm{sup 3} g{sup −1}. Good specific adsorption capacity for BPA was obtained in the sorption experiment and good class selectivity for BPA and its seven structural analogs (bisphenol F, bisphenol B, bisphenol E, bisphenol AF, bisphenol S, bisphenol AP and bisphenol Z) was demonstrated by the chromatographic evaluation experiment. The MIPMS as solid-phase extraction (SPE) packing material was then evaluated for extraction and clean-up of these bisphenols (BPs) from human urine samples. An accurate and sensitive analytical method based on the MIPMS-SPE coupled with HPLC-DAD has been successfully established for simultaneous determination of eight BPs from human urine samples with detection limits of 1.2–2.2 ng mL{sup −1}. The recoveries of BPs for urine samples at two spiking levels (100 and 500 ng mL{sup −1} for each BP) were in the range of 81.3–106.7% with RSD values below 8.3%.

Syngas fermentation by Clostridium carboxidivorans P7 in a horizontal rotating packed bed biofilm reactor with enhanced ethanol production

International Nuclear Information System (INIS)

Shen, Yanwen; Brown, Robert C.; Wen, Zhiyou

2017-01-01

Highlights: • A novel a horizontal rotating packed bed (h-RPB) reactor for syngas fermentation was reported. • The h-RPB reactor enhanced ethanol productivity by 3.3-folds compared to continuous stirred tank reactor (CSTR). • The h-RPB reactor has a unique feature of transfer gas from both bulk liquid phase and headspace phase. • The mass transfer in the headspace of h-PRB played an important role for enhanced ethanol production. - Abstract: Gasification of lignocellulosic biomass followed by syngas fermentation is a promising process for producing fuels and chemicals. Syngas fermentation, however, is commonly limited by low mass transfer rates. In this work, a horizontally oriented rotating packed bed (h-RPB) reactor was developed to improve mass transfer and enhance ethanol production. In the h-RPB reactor, cell attachment materials were packed in the reactor and half submerged in the liquid and half exposed to the headspace. With continuous rotation of the packing materials, the cells in biofilm were alternately in contact with liquid and headspace; thus, transport of syngas to the cells occurred in both the liquid phase and headspace. The volumetric mass transfer coefficient (k_La) of the h-RPB reactor was lower than that in a traditional continuous stirred tank reactor (CSTR), indicating the mass transfer in the liquid phase of h-PRB was lower than CSTR, and the mass transfer in the headspace phase played an important role in syngas fermentation. The syngas fermentation of Clostridium carboxidivorans P7 in h-RPB resulted in a 7.0 g/L titer and 6.7 g/L/day productivity of ethanol, respectively, 3.3 times higher than those obtained in a CSTR under the same operational conditions. The results demonstrate that the h-RPB reactor is an efficient system for syngas fermentation, making cellulosic ethanol biorefinery one step closer to technical and economic feasibility.

A study of wet deposition of atmospheric tritium releases at the Ontario Power Generation, Pickering Nuclear Generating Station

International Nuclear Information System (INIS)

Crooks, G.; DeWilde, J.; Yu, L.

2001-01-01

The Ontario Power Generation,Pickering Nuclear Generating Station (PNGS) has been investigating deposition of atmospheric releases of tritium on their site. This study has included numerical dispersion modelling studies conducted over the past three years, as well as an ongoing field monitoring study. The following paper will present results of the field monitoring study and make comparisons to the numerical modelling. The results of this study could be of potential use to nuclear stations in quantifying tritium deposition in near field regions where building wake effects dominate pollutant dispersion

Leakage rate from LOCA-aged inflatable airlock seals Pickering NGS 'B' personnel doors

International Nuclear Information System (INIS)

Fayle, G.W.; Cordingley, D.C.

1985-01-01

In order to demonstrate to the Atomic Energy Control Board that an air-lock inflatable seal will function after a LOCA exposure, an inflatable seal intended for personnel doors at the Pickering NGS 'B' was exposed to the thermal/moisture conditions of the LOCA requirement. While attending to determine the post-LOCA leakage rate it was found that additional leaks developed during each post-LOCA inflation/deflation cycle. The seal had been significantly and irreparably deteriorated by the LOCA exposure. The test has demonstrated that this type of LOCA exposed seal should not be expected to withstand either additional pressure above 207 kPa or additional inflation/deflation cycling. A higher inflation pressure and/or cycling will reduce the likelihood of a post-LOCA seal retaining an inflation pressure sufficient to prevent leakage across the seal

Evolution of CANDU vacuum building and pressure relief structures from Pickering NGS A to Darlington NGS A

International Nuclear Information System (INIS)

Beg, Z.M.; Ghosh, R.S.

1987-01-01

The vacuum building (VB) and pressure relief structures (PRS) are the unique features of multiple unit CANDU containments. In case of loss-of-coolant accident, the released radionuclides are drawn through the PRS into the subatmospheric VB, doused and contained without being released to the environment. This paper describes the differences in design, configuration and layout of the VB and PRS from Pickering NGS A to Darlington NGS A due to new developments in design concepts and to requirements which have proceeded from the experience gained in both the design and operation of the nuclear stations. (orig.)

Recent experience related to neutronic transients in Ontario Hydro CANDU nuclear generating stations

International Nuclear Information System (INIS)

Frescura, G.M.; Smith, A.J.; Lau, J.H.

1991-01-01

Ontario Hydro presently operates 18 CANDU reactors in the province of Ontario, Canada. All of these reactors are of the CANDU Pressurized Heavy Water design, although their design features differ somewhat reflecting the evolution that has taken place from 1971 when the first Pickering unit started operation to the present as the Darlington units are being placed in service. Over the last three years, two significant neutronic transients took place at the Pickering Nuclear Generating Station 'A' (NGS A) one of which resulted in a number of fuel failures. Both events provided valuable lessons in the areas of operational safety, fuel performance And accident analysis. The events and the lessons learned are discussed in this paper

IGORR 7: Proceedings of the 7. meeting of the International Group On Research Reactors

International Nuclear Information System (INIS)

1999-01-01

According to the subjects covered the papers presented at the meeting were divided into following sessions: New research reactor projects; secondary neutron sources; New research reactor facilities; Improvement of Research Reactors Facilities; Research and Development Needs

Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

International Nuclear Information System (INIS)

Loika, E.F.

1994-01-01

This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate

Interpretation of 131I hippuran renocystogram using vascular invasion segment systemic flow, and DYNA CAMERA II Picker

International Nuclear Information System (INIS)

Morcellet, J.L.; Baret, A.

A quantitative approximation of flows of fluids from each kidney (renal clearances urinary flows), of the hippuran mean stay time into each kidney was proposed. These times are decomposed into cortical transit mean time and into pyelocavities mean stay time. The use of a dual isotope scintillation Dyna Camera II Picker changes the collecting of the data and permits the simultaneous measurement of cardiac output which is required for their treatment. This treatment is carried out by the mean of a videotape recorder which authorizes delayed time work and by the mean of a hundred channels computer, which displays numerical data and their integration [fr

Light water reactor fuel analysis code FEMAXI-7. Model and structure

International Nuclear Information System (INIS)

Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki

2013-07-01

A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report describes in detail the design, basic theory and structure, models and numerical method of FEMAXI-7, and its improvements and extensions. (author)

Facile Route to Transparent, Strong, and Thermally Stable Nanocellulose/Polymer Nanocomposites from an Aqueous Pickering Emulsion.

Science.gov (United States)

Fujisawa, Shuji; Togawa, Eiji; Kuroda, Katsushi

2017-01-09

Cellulose nanofibril (CNF) is a promising nanofiller for polymer nanocomposite materials, and a critical challenge in designing these materials is organization of the nanostructure using a facile process. Here, we report a facile aqueous preparation process for nanostructured polystyrene (PS)/CNF composites via the formation of a CNF-stabilized Pickering emulsion. PS nanoparticles, with a narrow size distribution, were synthesized by free radical polymerization in water using CNF as a stabilizer. The nanoparticles were easily collected by filtration, and the resulting material had a composite structure of PS nanoparticles embedded in a CNF framework. The PS/CNF nanocomposite showed high optical transparency, strength, and thermal dimensional stability. Thus, this technique provides a simple and environmentally friendly method for the preparation of novel CNF/polymer nanocomposite materials.

Operating reactors licensing actions summary. Volume 5, No. 7

International Nuclear Information System (INIS)

1985-09-01

The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Light water reactor fuel analysis code FEMAXI-7; model and structure

International Nuclear Information System (INIS)

Suzuki, Motoe; Udagawa, Yutaka; Saitou, Hiroaki

2011-03-01

A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions. (author)

Fuel channel life limiting factors that dictate fuel channel maintenance requirements

International Nuclear Information System (INIS)

Richinson, P.J.; Wong, H.W.; Ellis, P.J.

1995-01-01

CANDU reactors have been operating for 33 years. The Nuclear Power Demonstration (NPD) Unit started up in 1962 and the prototype of CANDU, Douglas Point, started in 1967. The first commercial reactors, Pickering Units 1 and 2 both went into service in 1971 closely followed by Units 3 and 4 in 1972 and 1973 respectively. Operating commercial reactor experience represents over 10,000 pressure tubes, not including the replaced channels in all the Pickering A Units, and nearly 130,000 pressure tube operating years. No pressure tube has yet operated for its 30 year design lifetime of 210 KEFPH at 80% capacity factor. The longest operating time for pressure tubes to-date is about 120 KEFPH in Pickering Unit 4. Many lessons have been learned regarding pressure tube life limiting factors from the early CANDU units and these, together with the information obtained from an extensive pressure tube R and D program, have resulted in many design changes and improvements in material properties, mainly from manufacturing route changes. Reactors built recently are expected to achieve their 30 year design life. The development of Periodic and In-service Inspection programs and equipment, assessment methodologies and acceptance criteria, and the development of maintenance tooling and procedures are enabling the life limiting factors to be addressed in the currently operating units. The life limiting factors in currently operating Units are reviewed in relation to the experience gained from the early units, the R and D programs and the inspection and maintenance performed to date. (author)

THE INFLUENCE OF TRUST IN THE CONSTITUTION OF A BRAZILIAN COOPERATIVE OF SELECTIVE WASTE PICKERS

Directory of Open Access Journals (Sweden)

Dayanne Marciane Gonçalves

2016-03-01

Full Text Available Since the Brazilian public policy started to encourage solidarity economy in 2003, the number of projects and enterprises in this sector has steadily increased. Embeddedness has contributed to the understanding of organizational phenomena of solidarity economy and cooperatives. The aim of this study was to understand the influence of trust, from the perspective of Mark Granovetter’s social networks, on the constitution of a cooperative of urban recyclable waste pickers in southern Brazil between 1996 until early 2012, considered the foundation period. We used the qualitative method with a historical approach to social relationships and content analysis. Possible influences of trust were analysed based on the economic, social and political history of the cooperative. Among the main results, we highlight the existence of social relations before the constitution, defined by trust due to family identity and reputation built over time.

Project management lessons learned from building the Wendelstein 7-x stellerator fusion research reactor

Energy Technology Data Exchange (ETDEWEB)

Freire-Gormaly, M.; Gittens, A.; Zhang, L., E-mail: m.freire.gormaly@utoronto.ca, E-mail: antonio.gittens@mail.utoronto.ca, E-mail: lavender.zhang@outlook.com [University of Toronto, Toronto, ON (Canada)

2015-07-01

Wendelstein 7-X (W7-X) is the world's largest 'stellerator' nuclear fusion reactor being commissioned in Greifswald, Germany. It will inform the international fusion energy test device (ITER). The complexity of W7- X added challenges since industrial expertise to manufacture components did not exist. The construction was completed eight years behind schedule and almost 100% over budget. Key take-away lessons in project management were revealed from W7-X which can be applied to any nuclear project. These lessons are aligned with the project management knowledge areas of schedule, stakeholder, procurement, scope, schedule, cost, communication, risk, quality, human resources and procurement management. (author)

Project management lessons learned from building the Wendelstein 7-x stellerator fusion research reactor

International Nuclear Information System (INIS)

Freire-Gormaly, M.; Gittens, A.; Zhang, L.

2015-01-01

Wendelstein 7-X (W7-X) is the world's largest 'stellerator' nuclear fusion reactor being commissioned in Greifswald, Germany. It will inform the international fusion energy test device (ITER). The complexity of W7- X added challenges since industrial expertise to manufacture components did not exist. The construction was completed eight years behind schedule and almost 100% over budget. Key take-away lessons in project management were revealed from W7-X which can be applied to any nuclear project. These lessons are aligned with the project management knowledge areas of schedule, stakeholder, procurement, scope, schedule, cost, communication, risk, quality, human resources and procurement management. (author)

Design and construction of a 7,500 liter immobilized cell reactor-separator for ethanol production from whey

Energy Technology Data Exchange (ETDEWEB)

Dale, M.C.

1992-12-31

A 7,500 liter reactor/separator has been constructed for the production of ethanol from concentrated whey permeate. This unit is sited in Hopkinton IA, across the street from a whey generating cheese plant A two phase construction project consisting of (1) building and testing a reactor/separator with a solvent absorber in a single unified housing, and (2) building and testing an extractive distillation/product stripper for the recovery of anhydrous ethanol is under way. The design capacity of this unit is 250,000 gal/yr of anhydrous product. Design and construction details of the reactor/absorber separator are given, and design parameters for the extractive distillation system are described.

Digital reactor period meter type of NSSG-7

Energy Technology Data Exchange (ETDEWEB)

Glowacki, S W

1981-01-01

The paper presents the idea and electronic circuits of the Digital Reactor Period Meter. The instrument consists of a neutron ionisation chamber, the amplifier logarithming the output chamber current, the circuit taking two samples of the log amplifier output signal and subtracting them, the analog -to -digital dividing circuit and the scaler providing the final information of the reactor period value in seconds and in the digital form. Besides it, the instrument produces the acoustic signal in the case, when the rise-time of neutron flux exceeds the permitted value. The untypical construction of the reactor period meter has been developed to obtain both good measurement accuracy and the resistance against the electromagnetic background pulses interfering with the measuring process. The applied measuring system has been patented.

Fragger: a protein fragment picker for structural queries [version 2; referees: 2 approved

Directory of Open Access Journals (Sweden)

Francois Berenger

2018-04-01

Full Text Available Protein modeling and design activities often require querying the Protein Data Bank (PDB with a structural fragment, possibly containing gaps. For some applications, it is preferable to work on a specific subset of the PDB or with unpublished structures. These requirements, along with specific user needs, motivated the creation of a new software to manage and query 3D protein fragments. Fragger is a protein fragment picker that allows protein fragment databases to be created and queried. All fragment lengths are supported and any set of PDB files can be used to create a database. Fragger can efficiently search a fragment database with a query fragment and a distance threshold. Matching fragments are ranked by distance to the query. The query fragment can have structural gaps and the allowed amino acid sequences matching a query can be constrained via a regular expression of one-letter amino acid codes. Fragger also incorporates a tool to compute the backbone RMSD of one versus many fragments in high throughput. Fragger should be useful for protein design, loop grafting and related structural bioinformatics tasks.

PLUS7TM In-Reactor Operating Performance and Economics

International Nuclear Information System (INIS)

Kim, Kyutae; Jang, Youngki; Choi, Joonhyung; Lee, Jinseok; Kim, Yoonho; Suh, Jungmin

2006-01-01

KNFC has developed an advanced fuel, PLUS7 TM , for the Korean Standard Nuclear Power Plants(KSNPs) through the joint development program with Westinghouse. With the help of various out-of-pile tests, it is found that the PLUS7 TM shows much better performance than the current fuel, GUARDIAN TM from the safety and economy points of view. Now four Lead Test Assembles(LTAs) of the PLUS7 TM are being irradiated for the 3 rd cycle after the successful completion of the 1 st and 2 nd irradiation cycles. During the 1 st and 2 nd irradiation cycles, no fuel failure was observed at LTAs and their nuclear-related parameters matched their design values well. During the overhaul period, on the other hand, pool side examinations were performed for four LTAs to generate key in-reactor fuel performance data such as fuel rod and assembly growths, fuel rod-to-top nozzle gap, fuel assembly bow and twist, fuel rod bow, spacer grid width, fuel rod diameter and fuel rod oxide layer thickness. It is found that all measured values are bounded by upper and lower predicted ones. The detailed economic analyses have shown that significant fuel cycle cost can be reduced by more than one million dollars per cycle of one KSNP with the introduction of the PLUS7 TM assembly. Furthermore, more than one hundred million dollars with power up-rating of 5% can be saved annually for currently operating eight KSNPs, which is easily and safety achievable with the PLUS7 TM assembly

Steps to Advanced CANDU 600

International Nuclear Information System (INIS)

Oh, Yongshick; Brooks, G. L.

1988-01-01

The CANDU nuclear power system was developed from merging of AECL heavy water reactor technology with Ontario Hydro electrical power station expertise. The original four units of Ontario Hydro's Pickering Generating Station are the first full-scale commercial application of the CANDU system. AECL and Ontario Hydro then moved to the next evolutionary step, a more advanced larger scale design for four units at the Bruce Generating Station. CANDU 600 followed as a single unit nuclear electric power station design derived from an amalgam of features of the multiple unit Pickering and Bruce designs. The design of the CANDU 600 nuclear steam supply system is based on the Pickering design with improvements derived from the Bruce design. For example, most CANDU 600 auxiliary systems are based on Bruce systems, whereas the fuel handling system is based on the Pickering system. Four CANDU 600 units are in operation, and five are under construction in Romania. For the additional four units at Pickering Generating Station 'B', Ontario Hydro selected a replica of the Pickering 'A' design with limited design changes to maintain a high level of standardization across all eight units. Ontario Hydro applied a similar policy for the additional four units at Bruce Generating Station 'B'. For the four unit Darlington station, Ontario Hydro selected a design based on Bruce with improvements derived from operating experience, the CANDU 600 design and development programs

7. European conference of TRIGA reactor users. Conference papers

Energy Technology Data Exchange (ETDEWEB)

NONE

1982-07-01

At the Seventh European Conference of TRIGA Users, held in September 1982, in Istanbul, Turkey, the following aspects are discussed: safety aspects of TRIGA reactors; developments and improvements; operating and maintenance experiences; applications; reactor calculations; fuel cycle aspects and research programs.

7. European conference of TRIGA reactor users. Conference papers

International Nuclear Information System (INIS)

1982-01-01

At the Seventh European Conference of TRIGA Users, held in September 1982, in Istanbul, Turkey, the following aspects are discussed: safety aspects of TRIGA reactors; developments and improvements; operating and maintenance experiences; applications; reactor calculations; fuel cycle aspects and research programs

Glacial Lake Pickering: stratigraphy and chronology of a proglacial lake dammed by the North Sea Lobe of the British-Irish Ice Sheet

OpenAIRE

Evans, David J.A.; Bateman, Mark D.; Roberts, David H.; Medialdea, Alicia; Hayes, Laura; Duller, Geoff A.T.; Fabel, Derek; Clark, Chris D.

2016-01-01

We report the first chronology, using four new optically stimulated luminescence dates, on the sedimentary record of Glacial Lake Pickering, dammed by the North Sea Lobe of the British–Irish Ice Sheet during the Dimlington Stadial (24–11 ka cal BP). Dates range from 17.6 ± 1.0 to 15.8 ± 0.9 ka for the sedimentation of the Sherburn Sands at East Heslerton, which were formed by multiple coalescing alluvial fans prograding into the falling water levels of the lake and fed by progressively larger...

Design and installation of a strategically placed algae mesh barrier at OPG Pickering Nuclear Generating Station

International Nuclear Information System (INIS)

Marttila, D.; Patrick, P.; Gregoris, C.

2009-01-01

Ontario Power Generation's Pickering Nuclear has experienced a number of events in which attached algae have become entrained in the water intake costing approximately $30M over the 1995-2005 period as a result of deratings, Unit shutdowns and other operational issues. In 2005-2006 OPG and Kinectrics worked collaboratively on evaluating different potential solutions to reduce the impact of algae on the station. One of the solutions developed by Kinectrics included a strategically placed barrier net designed to regulate algae flow into the station intake. In 2006, Kinectrics designed and installed the system, the first of its kind at a Nuclear Power Plant in Canada. The system was operational by May 2007. OPG completed an effectiveness study in 2007 and concluded the barrier system had a beneficial effect on reducing algae impact on the station. (author)

Digital computer control on Canadian nuclear power plants -experience to date and the future outlook

International Nuclear Information System (INIS)

Pearson, A.

1977-10-01

This paper discusses the performance of the digital computer control system at Pickering through the years 1973 to 1976. This evaluation is based on a study of the Pickering Generating Station operating records. The paper goes on to explore future computer architectures and the advantages that could accrue from a distributed system approach. Also outlined are the steps being taken to develop these ideas further in the context of two Chalk River projects - REDNET, an advanced data acquisition system being installed to process information from engineering experiments in NRX and NRU reactors, and CRIP, a prototype communications network using cable television technology. (author)

Safety in nuclear power systems

International Nuclear Information System (INIS)

Myers, L.C.

1987-05-01

This paper discusses the issue of safety in complex energy systems and provides brief accounts of some of the most serious reactor accidents that have occurred to date. Details are also provided of Ontario Hydro's problems with Unit 2 at Pickering

Annual report/1979

International Nuclear Information System (INIS)

1980-04-01

Primary energy demand in Ontario in 1979 was up by 2.9 percent, compared to 2.7 percent in the previous year. Revised forecasts issued in January 1980 indicate Ontario's need for electricity is expected to grow by an average of 3.4 percent annually to the year 2000. Nuclear generation provided 29 percent of the total energy made available by Hydro, and Hydro's eight reactors at Pickering and Bruce continued to rank in the top 36 - four in the top 10 - when compared to the permance of 104 of the world's largest reactors. The provinical legislature's Select Committee on Hydro Affairs examined the safety of the CANDU system and concluded that is is 'acceptably safe'. Faced with reduced forecasts of electrical demands to the year 2000 the Board of Directors decided to stretch out the construction schedule of the Darlington station, to halt construction of the second half of the Bruce Heavy Water Plant D, and to complete but mothball the first half. Construction of Bruce Heavy Water Plant B was completed early in 1979. The A plant produced 599.8 megagrams of reactor-grade heavy water. A control room simulator for Bruce A nuclear generating station was ordered. Agreement was reached on rebuilding faulty boilers at Pickering B. A total of 757.6 megagrams of uranium was used to produce electrical energy and steam. Ontario Hydro continued involvement in uranium exploration. Studies on radioactive waste disposal are being carried out, with emphasis on interim storage and transportation. (LL)

Evaluation of severe accident risk in the Pickering a risk assessment

International Nuclear Information System (INIS)

Dinnie, K.S.; Raina, V.M.

1997-01-01

The nature of the design of commercial power plants is such that significant impacts on public health can only occur if a number of barriers fail. Rigorous design and licensing requirements ensure that the more likely accidents do not fail all these barriers and their contribution to risk is likely to be small. The task of estimating accident risk must, therefore, focus more towards those less likely but potentially more serious combinations of failures that are characterized by the following: a) a large release of fission products into the containment atmosphere, b) a breach in the containment envelope, and c) the existence of a driving force to expel the containment atmosphere to the outside environment. The likelihood of such conditions existing simultaneously during the course of an accident is expected to be small, such that experience and data regarding the behaviour of plant systems under such conditions is sparse or non-existent. The challenge of Probabilistic Safety Assessments (PSAs) is to examine the potential for severe accidents using approaches that are sufficiently detailed and realistic to provide valid information regarding plant risk and susceptibilities, while simple enough to keep the analysis manageable. This paper outlines the key features of the Pickering A Risk Assessment (PARA) (1) and the manner in which it addresses these issues, and provides some insights into the results and conclusions drawn from the study. (author)

Advances in Reactor Physics, Mathematics and Computation. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.

Reactor of the XXI century

International Nuclear Information System (INIS)

Zhotabaev, Zh.R.; Solov'ev, Yu.A.

2001-01-01

The advantages of both molten salt reactors (MSR) and homogenous molten salt reactors (HMSR) are illuminated. It is noted that the MSR possess accident probability A=10 -6 1/reactor.years and the HMSR with integral configuration has A=10 -7 1/reactor.years. The methods for these reactors technological problems solution - tritium removal, salt melt circulation and capacity pick up - are discussed

The case for new nuclear

International Nuclear Information System (INIS)

Luxat, J.C.

2013-01-01

Over a 22 year period from 1971 to 1993 a total of 20 reactor units were brought into service - an average of approximately one unit per year. Ontario Hydro constructed the four-unit Pickering A station, four units at Bruce A, four units at Pickering B, four units at Bruce B and four units at Darlington during this period. This represents a capacity of nearly 14,000 MW, as shown in Figure 1. During this period there was a large increase in industrial capacity in Ontario, particularly in manufacturing, driven in large measure by the incentives offered by low electricity prices, skilled workers and a good health care system. Subsequently in the mid-1990's the Pickering A and Bruce A units were laid up and maintenance efforts were focused on the Pickering B, Bruce B and Darlington stations. Two of the four units at Pickering A were returned to service in the early 2000's and the four units of Bruce A were returned to service with two units being refurbished. By 2010 nuclear capacity in the province had returned to 12,800 MW. The Ontario Long Term Energy Plan (LTEP) announced at the beginning of December does not include new build nuclear but does include refurbishment of the Darlington station as well as two units at Bruce A and four units at Bruce B. The six units at Pickering will be shut down by 2020. As shown in Figure 1, this will reduce the nuclear capacity from the current 12,800 MW to 8000 MW when the Pickering A and B units are removed from service in 2020 and the refurbishment of Darlington and Bruce units proceeds starting in 2016 and projected to complete by 2031. This will be the lowest nuclear generating capacity in the province since 1985. (author)

Suitability of CR-39 dosimeters for personal dosimetry around CANDU reactors

International Nuclear Information System (INIS)

Cross, W.G.

1992-08-01

The capabilities and limitations of CR-39 damage track detectors have been evaluated for their use as personal neutron dosimeters around CANDU reactors. Since the energy response is a critical characteristic, the neutron energy spectra expected within CANDU containments were studied. In the boiler rooms, around the moderator cooling systems, and in most of the fueling machine vaults, the spectra vary considerably, but the majority of the dose is expected to be delivered by neutrons above 80 keV, the approximate threshold for electrochemically-etched CR-39 detectors. In the Pickering A fueling machine vault, and in areas in other stations to which neutrons from reactors have been multiply scattered, lower energy neutrons may be important. In nearly all areas where people work, it appears that working times will be limited by gamma rays rather than by neutrons. The characteristics of other neutron dosimeters - bubble and superheated drop detectors, albedo detectors, and Si real-time detectors - were also reviewed. For workers who typically receive neutron doses that are small compared with regulatory limits, CR-39 is the most suitable available dosimeter for demonstrating compliance. All single dosimeters have poor angular response over the range 0 to 180 degrees because of the shielding of the body. Albedo and Si detectors have particularly poor energy responses over the energy range of importance. Bubble and superheated drop detectors have the advantages of immediate readout and high sensitivity, but the disadvantages of inability to integrate doses over a long period, temperature dependence, very limited range and higher cost. (Author) (110 refs., 45 figs.)

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

Energy Technology Data Exchange (ETDEWEB)

Block, R.C.; Feiner, F. [American Nuclear Society, La Grange Park, IL (United States)

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers deal with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

International Nuclear Information System (INIS)

Block, R.C.; Feiner, F.

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers deal with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately

Oxidation and deuterium uptake of Zr-2.5Nb pressure tubes in CANDU-PHW reactors

International Nuclear Information System (INIS)

Urbanic, V.F.; Warr, B.D.; Manolescu, A.; Chow, C.K.; Shanahan, M.W.

1989-01-01

Oxidation and deuterium uptake in Zr-2.5Nb pressure tubes are being monitored by destructive examination of tubes removed from commercial Canadian deuterium uranium pressurized heavy-water (CANDU-PHW) stations and by analyses of microsamples, obtained in-situ, from the inside surface of tubes in the reactor. Unlike Zircaloy-2, there is no evidence for any acceleration in the oxidation rate for exposures up to about 4500 effective full power days. Changes towards a more equilibrium microstructure during irradiation may be partly responsible for maintaining the low oxidation rate, since thermal aging treatments, producing similar microstructural changes in initially cold worked tubes, were found to improve out-reactor corrosion resistance in 589 K water. With one exception, the deuterium uptake in Zr-2.5Nb tubes has been remarkably low and no greater than 3-mg/kg deuterium per year (0.39 mg/dm 2 hydrogen per year) . The exception is the most recent surveillance tube removed from Pickering (NGS) Unit 3, which had a deuterium content near the outlet end about five times higher than that seen in the previous tube examined. Current investigations suggest that most of the uptake in that tube may have come from the gas annulus surrounding the tube where deuterium exists as an impurity, and oxidation has been insufficient to maintain a protective oxide film. Results from weight gain measurements, chemical analyses, metallography, scanning electron microscopy, and transmission electron microscopy of irradiated pressure tubes and of small coupons exposed out reactor are presented and discussed with respect to the observed corrosion and hydriding behavior of CANDU-PHW pressure tubes. (author)

The examination of the ruptured Zircaloy-2 pressure tube from Pickering NGS Unit 2

International Nuclear Information System (INIS)

Cheadle, B.A.; Smith, A.D.; Baskin, C.C.

1985-07-01

On 1983 August 01 a Zircaloy-2 pressure tube in Pickering NGS Unit 2 ruptured. All the fuel channel components, the fuel bundles, pressure tube, end fittings, garter springs and calandria tubes were shipped to Chalk River Nuclear Laboratories for examination to determine the cause of the rupture. The examination showed that the rupture initiated at a series of hydride blisters on the outside surface of the pressure tube. The blisters formed because of the garter spring spacers between the pressure tube and calandria tube was about one metre out of position. This allowed the horizontal pressure tube to sag by creep and touch the cool calandria tube. The resulting thermal gradients in the pressure tube concentrated the hydrogen and deuterium at the cool zones and blisters of solid hydride formed. Cracks initiated at several of the blisters and linked together to form a partial through wall critical crack which initiated the final rupture. The video presentation shows how the examination of the fuel channel components was conducted in underwater bays and shielded cells and explains the sequence of events that caused the rupture

Production of Concentrated Pickering Emulsions with Narrow Size Distributions Using Stirred Cell Membrane Emulsification.

Science.gov (United States)

Manga, Mohamed S; York, David W

2017-09-12

Stirred cell membrane emulsification (SCME) has been employed to prepare concentrated Pickering oil in water emulsions solely stabilized by fumed silica nanoparticles. The optimal conditions under which highly stable and low-polydispersity concentrated emulsions using the SCME approach are highlighted. Optimization of the oil flux rates and the paddle stirrer speeds are critical to achieving control over the droplet size and size distribution. Investigating the influence of oil volume fraction highlights the criticality of the initial particle loading in the continuous phase on the final droplet size and polydispersity. At a particle loading of 4 wt %, both the droplet size and polydispersity increase with increasing of the oil volume fraction above 50%. As more interfacial area is produced, the number of particles available in the continuous phase diminishes, and coincidently a reduction in the kinetics of particle adsorption to the interface resulting in larger polydisperse droplets occurs. Increasing the particle loading to 10 wt % leads to significant improvements in both size and polydispersity with oil volume fractions as high as 70% produced with coefficient of variation values as low as ∼30% compared to ∼75% using conventional homogenization techniques.

Electrospun composite matrices of poly(ε-caprolactone)-montmorillonite made using tenside free Pickering emulsions.

Science.gov (United States)

Samanta, Archana; Takkar, Sonam; Kulshreshtha, Ritu; Nandan, Bhanu; Srivastava, Rajiv K

2016-12-01

The production of composite electrospun matrices of poly(ε-caprolactone) (PCL) using an emulsifier-free emulsion, made with minimal organic solvent, as precursor is reported. Pickering emulsions of PCL were prepared using modified montmorillonite (MMT) clay as the stabilizer. Hydrophobic tallow group of the modified MMT clay resulted in analogous interaction of clay with oil and aqueous phase and its adsorption at the interface to provide stability to the resultant emulsion. Composite fibrous matrices of PCL and MMT were produced using electrospinning under controlled conditions. The fiber fineness was found to alter with PCL concentration and volume fraction of the aqueous and oil phases. A higher tensile strength and modulus was obtained with inclusion of MMT in PCL electrospun matrix in comparison to a matrix made using neat PCL. The presence of clay in the fibrous matrix did not change the cell proliferation efficiency in comparison to neat PCL matrix. Composite fibrous matrices of PCL/MMT bearing enhanced tensile properties may find applications in areas other than tissue engineering for example food packaging and filtration. Copyright © 2016 Elsevier B.V. All rights reserved.

International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

Investigations of titamium and zirconium hydrides to determine suitability of recoverable tritium immobilization for the Pickering tritium removal system

International Nuclear Information System (INIS)

Noga, J.O.

1981-11-01

A tritium removal system will be constructed at Pickering Nuclear Generating station to reduce the adverse effects of this radioactive hydrogen isotope. This report summarizes various properties of titanium and zirconium sponge hydrides which have been selected as suitable candidates for tritium product immobilization. Equilibrium pressure-composition-temperature data indicates that both materials behave suitably to provide a safe, solid form of tritium storage. Titanium tritide is recommended as the best choice due to higher dissociation pressures which can be achieved at equivalent temperatures when compared to zirconium tritide. Higher dissociation pressures would result in faster and more efficient recovery of tritium gas from the immobilized state. It is evident from the stability of these compounds that their utilization as tritides will greatly enhance the integrity of tritium storage

Impact of proposed research reactor standards on reactor operation

Energy Technology Data Exchange (ETDEWEB)

Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

1974-07-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

Impact of proposed research reactor standards on reactor operation

International Nuclear Information System (INIS)

Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

1974-01-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

Energy Technology Data Exchange (ETDEWEB)

Shoushtari, M.K.; Kakavand, T. [Faculty of Science, University of Zanjan, Zanjan, P.O. Box 451-313 (Iran, Islamic Republic of); Sadat Kiai, S.M., E-mail: sadatkiai@yahoo.co [Nuclear Science and Technology Research Institute (NSTR), Nuclear Science Research, A.E.O.I., P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of); Ghaforian, H. [Faculty of Science and Technology of Marine, Tehran (Iran, Islamic Republic of)

2010-03-01

The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity (rho), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.

Fabrication and evaluation of chitosan/NaYF4:Yb3+/Tm3+ upconversion nanoparticles composite beads based on the gelling of Pickering emulsion droplets

International Nuclear Information System (INIS)

Yan, Huiqiong; Chen, Xiuqiong; Shi, Jia; Shi, Zaifeng; Sun, Wei; Lin, Qiang; Wang, Xianghui; Dai, Zihao

2017-01-01

The rare earth ion doped upconversion nanoparticles (UCNPs) synthesized by hydrophobic organic ligands possess poor solubility and low fluorescence quantum yield in aqueous media. To conquer this issue, NaYF 4 :Yb 3+ /Tm 3+ UCNPs, synthesized by a hydrothermal method, were coated with F127 and then assembled with chitosan to fabricate the chitosan/NaYF 4 :Yb 3+ /Tm 3+ composite beads (CS/NaYF 4 :Yb 3+ /Tm 3+ CBs) by Pickering emulsion system. The characterization results revealed that the as-synthesized NaYF 4 :Yb 3+ /Tm 3+ UCNPs with an average size of 20 nm exhibited spherical morphology, high crystallinity and characteristic emission upconversion fluorescence with an overall blue color output. The NaYF 4 :Yb 3+ /Tm 3+ UCNPs were successfully conjugated on the surface of chitosan beads by the gelling of emulsion droplets. The resultant CS/NaYF 4 :Yb 3+ /Tm 3+ CBs showed good upconversion luminescent property, drug-loading capacity, release performance and excellent biocompatibility, exhibiting great potentials in targeted drug delivery and tissue engineering with potential tracking capability and lasting release performance. - Highlights: • NaYF 4 :Yb 3+ /Tm 3+ UCNPs were coated by F127 to improve aqueous dispersibility. • NaYF 4 :Yb 3+ /Tm 3+ UCNPs were assembled with chitosan to fabricate the composite beads (CMs). • Pickering emulsions stabilized by UCNPs exhibited uniform and satisfactory emulsion droplets. • The CMs prepared by the gelling of emulsion droplet preserved upconversion luminescent property. • The resultant CMs showed good drug-loading capacity, release performance and biocompatibility.

Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

Energy Technology Data Exchange (ETDEWEB)

Barnett, P. C. [Canadian General Electric (Canada)

1968-07-15

The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

Validation of moderator-level reactivity coefficient using station data

Energy Technology Data Exchange (ETDEWEB)

Younis, M.; Martchouk, I., E-mail: mohamed.younis@amecfw.com, E-mail: iouri.martchouk@amecfw.com [Amec Foster Wheeler, Toronto, ON (Canada); Buchan, P.D., E-mail: david.buchan@opg.com [Ontario Power Generation, Pickering, ON (Canada)

2015-07-01

The reactivity effect due to variations in the moderator level has been recognized as a reactor physics phenomenon of importance during normal operation and accident analysis. The moderator-level reactivity coefficient is an important parameter in safety analysis of CANDU reactors, e.g., during Loss of Moderator Heat Sink as well as in the simulation of Reactor Regulating System action in CANDU reactors that use moderator level for reactivity control. This paper presents the results of the validation exercise of the reactor-physics toolset using the measurements performed in Pickering Unit 4 in 2003. The capability of the code suite of predicting moderator-level reactivity effect was tested by comparing measured and predicted reactor-physics parameters. (author)

Preparation of Lignin/Sodium Dodecyl Sulfate Composite Nanoparticles and Their Application in Pickering Emulsion Template-Based Microencapsulation.

Science.gov (United States)

Pang, Yuxia; Wang, Shengwen; Qiu, Xueqing; Luo, Yanling; Lou, Hongming; Huang, Jinhao

2017-12-20

Lignin is a vastly underutilized biomass resource. The preparation of water-dispersed lignin nanoparticles is an effective way to realize the high-value utilization of lignin. However, the currently reported preparation methods of lignin nanoparticles still have some drawbacks, such as the requirement for toxic organic solvent or chemical modification, complicated operation process, and poor dispersibility. Here, lignin/sodium dodecyl sulfate (SDS) composite nanoparticles (LSNPs) with outstanding water dispersibility and a size range of 70-200 nm were facilely prepared via acidifying the mixed basic solution of alkaline lignin and SDS. No harsh chemical was needed. The formation mechanism was systematically studied. Results indicated that the LSNPs were obtained by acid precipitation of the mixed micelles formed by the self-assembly of lignin and SDS. In addition, on the basis of the LSNP-stabilized Pickering emulsions, lignin/polyurea composite microcapsules combining the excellent chemical stability of a synthetic polyurea shell with the fantastic antiphotolysis and antioxidant properties of lignin were successfully prepared.

Automated phase picker and source location algorithm for local distances using a single three component seismic station

International Nuclear Information System (INIS)

Saari, J.

1989-12-01

The paper describes procedures for automatic location of local events by using single-site, three-component (3c) seismogram records. Epicentral distance is determined from the time difference between P- and S-onsets. For onset time estimates a special phase picker algorithm is introduced. Onset detection is accomplished by comparing short-term average with long-term average after multiplication of north, east and vertical components of recording. For epicentral distances up to 100 km, errors seldom exceed 5 km. The slowness vector, essentially the azimuth, is estimated independently by using the Christoffersson et al. (1988) 'polarization' technique, although a priori knowledge of the P-onset time gives the best results. Differences between 'true' and observed azimuths are generally less than 12 deg C. Practical examples are given by demonstrating the viability of the procedures for automated 3c seismogram analysis. The results obtained compare favourably with those achieved by a miniarray of three stations. (orig.)

Compact stellarators as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

2001-01-01

Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

Compact torsatron reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

1988-05-01

Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R 0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R 0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

Static yield stress of a magnetorheological fluid containing Pickering emulsion polymerized Fe2O3/polystyrene composite particles.

Science.gov (United States)

Seo, Youngwook P; Kwak, Soonjong; Choi, Hyoung Jin; Seo, Yongsok

2016-02-01

The flow behaviors of magnetorheological (MR) suspensions containing Pickering emulsion polymerized Fe2O3/polystyrene (PS) composite particles were reanalyzed using the Seo-Seo model. The experimental shear stress data obtained experimentally from the magnetorheological fluid fit well to the Seo-Seo model, indicating that this model can describe the structural reformation process of the aligned fibers at various shear rates. Unlike the dynamic yield stress obtained from the Cho-Choi-Jhon (CCJ) model, the static yield stresses obtained from the Seo-Seo model exhibit the same quadratic dependence on the magnetic field strength for both pure Fe2O3 particle suspension and Fe2O3/PS particle suspensions, which is in agreement with the predictions of the polarization model. The static yield stress plausibly explains the difference in underlying mechanism of MR fluids. Copyright © 2015 Elsevier Inc. All rights reserved.

Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor

International Nuclear Information System (INIS)

Raisic, N.

1962-09-01

In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined

In situ sampling for pressure tube deuterium concentration

International Nuclear Information System (INIS)

Harrington, A.J.; Kittmer, C.A.

1988-01-01

The present method of assessing the useful life of pressure tubes in CANDU (CANada Deuterium Uranium) reactors requires the periodic removal and examination of a tube. Special tooling was developed at Atomic Energy of Canada Limited (AECL) to obtain a sample of material from a pressure tube without removing the tube from the reactor. The sampling tool concept has been successfully used by Ontario Hydro during scheduled outages at the Pickering Nuclear Generating Station (PNGS). (author)

Advances in Reactor Physics, Mathematics and Computation. Volume 1

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume one, are divided into 6 sessions bearing on: - session 1: Advances in computational methods including utilization of parallel processing and vectorization (7 conferences) - session 2: Fast, epithermal, reactor physics, calculation, versus measurements (9 conferences) - session 3: New fast and thermal reactor designs (9 conferences) - session 4: Thermal radiation and charged particles transport (7 conferences) - session 5: Super computers (7 conferences) - session 6: Thermal reactor design, validation and operating experience (8 conferences).

Input/output manual of light water reactor fuel analysis code FEMAXI-7 and its related codes

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Saitou, Hiroaki [ITOCHU Techno-Solutions Corporation, Tokyo (Japan)

2013-10-15

A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (author)

Input/output manual of light water reactor fuel analysis code FEMAXI-7 and its related codes

International Nuclear Information System (INIS)

Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki

2013-10-01

A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (author)

Multipurpose research reactors

International Nuclear Information System (INIS)

1988-01-01

The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

Commissioning of research reactors. Safety guide

International Nuclear Information System (INIS)

2006-01-01

The objective of this Safety Guide is to provide recommendations on meeting the requirements for the commissioning of research reactors on the basis of international best practices. Specifically, it provides recommendations on fulfilling the requirements established in paras 6.44 and 7.42-7.50 of International Atomic Energy Agency, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005) and guidance and specific and consequential recommendations relating to the recommendations presented in paras 615-621 of International Atomic Energy Agency, Safety in the Utilization and Modification of Research Reactors, Safety Series No. 35-G2, IAEA, Vienna (1994) and paras 228-229 of International Atomic Energy Agency, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series No. 35-G1, IAEA, Vienna (1994). This Safety Guide is intended for use by all organizations involved in commissioning for a research reactor, including the operating organization, the regulatory body and other organizations involved in the research reactor project

Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

International Nuclear Information System (INIS)

Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

2007-01-01

Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at ∼2.4, ∼7 and ∼11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of ∼7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10 15 n/cm 2 /s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between ∼410 deg. C and ∼645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

Electrosleeve process for in-situ nuclear steam generator repair

International Nuclear Information System (INIS)

Renaud, E.; Brennenstuhl, A.M.; Stewart, D.R.; Gonzalez, F.

2000-01-01

Degradation of steam generator tubing by localized corrosion is a widespread problem in the nuclear industry that can lead to costly forced outages, unit derating, steam generator replacement or even the permanent shutdown of a reactor. In response to the onset of steam generator degradation at Ontario Power Generation's Pickering Nuclear Generating Station (PNGS) Unit 5, and the determined unsuitability of conventional repair methods (mechanically expanded or welded sleeves) for Alloy 400, an alternative repair technology was developed. Electrosleeve is a non-intrusive, low-temperature process that involves the electrodeposition of a nanocrystalline nickel microalloy forming a continuously bonded, structural layer over the internal diameter of the degraded region. This technology is designed to provide a long-term pressure boundary repair, fully restoring the structural integrity of the damaged region to its original state. This paper describes the Electrosleeve process for steam generator tubing repair and the unique properties of the advanced sleeve material. The successful installation of fourteen Electrosleeves that have been in service for more than six years in Alloy 400 tubing at the Pickering-S CANDU unit, and the more recent (Nov. 99) extension of the technology to Alloy 600 by the installation of 57 sleeves in a U.S. pressurized water reactor (PWR) at Callaway, is presented. The Electrosleeve process has been granted a conditional license by the U.S. Nuclear Regulatory Commission (NRC). In Canada, the process of licensing Electrosleeve with the CNSC / TSSA has begun. (author)

Testing of a 7-tube palladium membrane reactor for potential use in TEP

International Nuclear Information System (INIS)

Carlson, Bryan J.; Trujillo, Stephen; Willms, R. Scott

2010-01-01

A Palladium Membrane Reactor (PMR) consists of a palladium/silver membrane permeator filled with catalyst (catalyst may be inside or outside the membrane tubes). The PMR is designed to recover tritium from the methane, water, and other impurities present in fusion reactor effluent. A key feature of a PMR is that the total hydrogen isotope content of a stream is significantly reduced as (1) methane-steam reforming and/or water-gas shift reactions proceed on the catalyst bed and (2) hydrogen isotopes are removed via permeation through the membrane. With a PMR design matched to processing requirements, nearly complete hydrogen isotope removals can be achieved. A 3-tube PMR study was recently completed. From the results presented in this study, it was possible to conclude that a PMR is appropriate for TEP, perforated metal tube protectors function well, platinum on aluminum (PtA) catalyst performs the best, conditioning with air is probably required to properly condition the Pd/Ag tubes, and that CO/CO 2 ratios maybe an indicator of coking. The 3-tube PMR had a permeator membrane area of 0.0247 m 2 and a catalyst volume to membrane area ratio of 4.63 cc/cm 2 (with the catalyst on the outside of the membrane tubes and the catalyst only covering the membrane tube length). A PMR for TEP will require a larger membrane area (perhaps 0.35 m 2 ). With this in mind, an intermediate sized PMR was constructed. This PMR has 7 permeator tubes and a total membrane area of 0.0851 m 2 . The catalyst volume to membrane area ratio for the 7-tube PMR was 5.18 cc/cm 2 . The total membrane area of the 7-tube PMR (0.0851 m 2 ) is 3.45 times larger than total membrane area of the 3-tube PMR (0.0247 m 2 ). The following objectives were identified for the 7-tube PMR tests: (1) Refine test measurements, especially humidity and flow; (2) Refine maintenance procedures for Pd/Ag tube conditioning; (3) Evaluate baseline PMR operating conditions; (4) Determine PMR scaling method; (5) Evaluate PMR

Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 16, 2011, 7:00 PM status

International Nuclear Information System (INIS)

2011-01-01

This situation note is established according to the information gained on March 16, 2011, at 7:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of all 6 reactors of the Fukushima I site (Dai-ichi) and of their spent fuel pools, as well as the situation of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

Input/output manual of light water reactor fuel performance code FEMAXI-7 and its related codes

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Saitou, Hiroaki [ITOCHU Techno-Solutions Corp., Tokyo (Japan)

2012-07-15

A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which has been fully disclosed in the code model description published recently as JAEA-Data/Code 2010-035. The present manual, which is the counterpart of this description, gives detailed explanations of operation method of FEMAXI-7 code and its related codes, methods of Input/Output, methods of source code modification, features of subroutine modules, and internal variables in a specific manner in order to facilitate users to perform a fuel analysis with FEMAXI-7. This report includes some descriptions which are modified from the original contents of JAEA-Data/Code 2010-035. A CD-ROM is attached as an appendix. (author)

Input/output manual of light water reactor fuel performance code FEMAXI-7 and its related codes

International Nuclear Information System (INIS)

Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki

2012-07-01

A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which has been fully disclosed in the code model description published recently as JAEA-Data/Code 2010-035. The present manual, which is the counterpart of this description, gives detailed explanations of operation method of FEMAXI-7 code and its related codes, methods of Input/Output, methods of source code modification, features of subroutine modules, and internal variables in a specific manner in order to facilitate users to perform a fuel analysis with FEMAXI-7. This report includes some descriptions which are modified from the original contents of JAEA-Data/Code 2010-035. A CD-ROM is attached as an appendix. (author)

Fusion reactor materials

International Nuclear Information System (INIS)

Anon.

1977-01-01

The following topics are briefly discussed: (1) surface blistering studies on fusion reactor materials, (2) TFTR design support activities, (3) analysis of samples bombarded in-situ in PLT, (4) chemical sputtering effects, (5) modeling of surface behavior, (6) ion migration in glow discharge tube cathodes, (7) alloy development for irradiation performance, (8) dosimetry and damage analysis, and (9) development of tritium migration in fusion devices and reactors

Performance of Canadian commercial nuclear units and heavy water plants

International Nuclear Information System (INIS)

Woodhead, L.W.; Ingolfsrud, L.J.

The operating history of Canadian commercial CANDU type reactors, i.e. Pickering generating station-A, is described. Capacity factors and unit energy costs are analyzed in detail. Equipment performance highlights are given. The performance of the two Canadian heavy water plants is described and five more are under construction or planned. (E.C.B.)

European commission - 7th framework programme. The collaborative project on European sodium fast reactor (CP ESFR)

International Nuclear Information System (INIS)

Fiorini, G.L.

2009-01-01

The paper summarizes the key characteristics of the four years large Collaborative Project on European Sodium Fast Reactor (CP ESFR - 2009-2012); the CP ESFR follows the 6th FP project named 'Roadmap for a European Innovative SOdium cooled FAst Reactor - EISOFAR' further identifying, organizing and implementing a significant part of the needed R and D effort. The CP ESFR merges the contribution of 25 european partners; it will be realized under the aegis of the 7th FP under the Area - Advanced Nuclear Systems with a refund from the European Commission of 5.8 M euro (11.55 M euro total budget). It will be a key component of the European Sustainable Nuclear Energy Technology Platform (SNE TP) and its Strategic Research Agenda (SRA). The inputs for the project are the key research goals for fourth generation of European sodium cooled fast reactors which can be summarized as follow: an improved safety with in particular the achievement of a robust architecture vis a vis of abnormal situations and the robustness of the safety demonstrations; the guarantee of a financial risk comparable to that of the other means of energy production; a flexible and robust management of the nuclear materials and especially the waste reduction through the Minor Actinides burning. (author)

Production of starch nanoparticles by dissolution and non-solvent precipitation for use in food-grade Pickering emulsions.

Science.gov (United States)

Saari, Hisfazilah; Fuentes, Catalina; Sjöö, Malin; Rayner, Marilyn; Wahlgren, Marie

2017-02-10

The aim of this study was to investigate non-solvent precipitation of starch to produce nanoparticles that could be used in Pickering emulsions. The material used was waxy maize, modified with octenyl succinic anhydride. Different methods of non-solvent precipitation were investigated, and a method based on direct mixing of an 8% starch solution and ethanol (ratio 1:1) was found to produce the smallest particles. The particle size was measured using AFM and AF4, and was found to be in the range 100-200nm. However, both larger particles and aggregates of nanoparticles were observed. The emulsion produced using the precipitated starch particles had a droplet size that between 0.5 and 45μm, compared to emulsions produced from waxy maize granules, in which had a size of 10-100μm. The drop in size contributed to increased stability against creaming. The amount of starch used for emulsion stabilization could also be substantially reduced. Copyright © 2016 The Author(s). Published by Elsevier Ltd.. All rights reserved.

Self-Assembled Colloidal Particle Clusters from In Situ Pickering-Like Emulsion Polymerization via Single Electron Transfer Mechanism.

Science.gov (United States)

Yuan, Jinfeng; Zhao, Weiting; Pan, Mingwang; Zhu, Lei

2016-08-01

A simple route is reported to synthesize colloidal particle clusters (CPCs) from self-assembly of in situ poly(vinylidene fluoride)/poly(styrene-co-tert-butyl acrylate) [PVDF/P(St-co-tBA)] Janus particles through one-pot seeded emulsion single electron transfer radical polymerization. In the in situ Pickering-like emulsion polymerization, the tBA/St/PVDF feed ratio and polymerization temperature are important for the formation of well-defined CPCs. When the tBA/St/PVDF feed ratio is 0.75 g/2.5 g/0.5 g and the reaction temperature is 35 °C, relatively uniform raspberry-like CPCs are obtained. The hydrophobicity of the P(St-co-tBA) domains and the affinity of PVDF to the aqueous environment are considered to be the driving force for the self-assembly of the in situ formed PVDF/P(St-co-tBA) Janus particles. The resultant raspberry-like CPCs with PVDF particles protruding outward may be promising for superhydrophobic smart coatings. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Neutronics of a mixed-flow gas-core reactor

International Nuclear Information System (INIS)

Soran, P.D.; Hansen, G.E.

1977-11-01

The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF 6 (either U-233 or U-235) and various parameters were varied. A four-cell reactor is not practical nor is the U-235 fueled seven-cell radial reactor; however, the 7-cell U-233 radial and scallop reactors can satisfy all design criteria. The mixed flow gas core reactor is a very attractive reactor concept and warrants further investigation

Light water reactor fuel analysis code FEMAXI-7. Model and structure [Revised edition

International Nuclear Information System (INIS)

Suzuki, Motoe; Udagawa, Yutaka; Amaya, Masaki; Saitou, Hiroaki

2014-03-01

A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report is the revised edition of the first one which describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions. The first edition, JAEA-Data/Code 2010-035, was published in 2010. The first edition was extended by orderly addition and disposition of explanations of models and organized as the revised edition after three years interval. (author)

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Martinc, R.; Cupac, S.; Stanic, A.

1990-01-01

RA reactor was not operated during the past five years due to the renewal and reconstruction of the reactor systems, which in underway. In the period from 1986-1990, reactor was operated only 144 MWh in 1986, for the need of testing the reactor systems and possibility of irradiating 125 I. Reactor will not be operated in 1991 because of the exchange of complete instrumentation which is planned to be finished by the end of 1991. It is expected to start operation in May 1992. That is why this annex includes the plan of reactor operation for period of nine months starting from from the moment of start-up. It is planned to operate the reactor at 0.02 MW power first three months, to increase the power gradually and reach 3.5 MW after 8 months of operation. It is foreseen to operate the reactor at 4.7 MW from the tenth month on [sr

Compact power reactor

International Nuclear Information System (INIS)

Wetch, J.R.; Dieckamp, H.M.; Wilson, L.A.

1978-01-01

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector

Fuel channel design improvements for large CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Villamagna, A; Price, E G; Field, G J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

From the initial designs used in NPD and Douglas point reactors, the CANDU fuel channel and its components have undergone considerable development. Two major designs have evolved: the Pickering/CANDU 6 design which has 12 fuel bundles in the core and where the new fuel is inserted into the inlet end, and the Bruce/Darlington design which has 13 bundles in the channel and where new fuel is inserted into the outlet end. In the development of a single unit CANDU reactor of the size of a Bruce or Darlington unit which would use a Darlington design calandria, the decision has been made to use the CANDU 6 fuel channel rather than the Darlington design. The CANDU 6 channel has provided excellent performance and will not encounter the degree of maintenance required for the Bruce/Darlington design. The channel design in turn influences the fuelling machine/fuel handling concepts required. The changes to the CANDU 6 fuel channel design to incorporate it in the large unit are small. In fact, the changes that are proposed relate to the desire to increase margins between pressure tube properties and design conditions or ameliorate the consequences of postulated accident conditions, rather than necessary adaptation to the larger unit. Better properties have been achieved in the pressure tube material resulting from alloy development program over the past 10 years. Pressure tubes can now he made with very low hydrogen concentrations so that the hydrogen picked up as deuterium will not exceed the terminal solid solubility for the in-core region in 30 years. The improvements in metal chemistry allow the production of high toughness tubes that retain a high level of toughness during service. A small increase in wall thickness will reduce the dimensional changes without significantly affecting burnup. Changes to increase safety margins from postulated accidents are concentrated on containing the consequences of pressure tube damage. The changes are concentrated on the calandria tube

Molten salt reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Simon, N.; Renault, C.

2014-01-01

Molten salt reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. The principle of this reactor is very innovative: the nuclear fuel is dissolved in the coolant which allows the online reprocessing of the fuel and the online recovery of the fission products. A small prototype: the Molten Salt Reactor Experiment (MSRE - 8 MWt) was operating a few years in the sixties in the USA. The passage towards a fast reactor by the suppression of the graphite moderator leads to the concept of Molten Salt Fast Reactor (MSFR) which is presently studied through different European projects such as MOST, ALISIA and EVOL. Worldwide the main topics of research are: the adequate materials resisting to the high level of corrosiveness of the molten salts, fuel salt reprocessing, the 3-side coupling between neutron transport, thermohydraulics and thermo-chemistry, the management of the changing chemical composition of the salt, the enrichment of lithium with Li 7 in the case of the use of lithium fluoride salt and the use of MSFR using U 233 fuel (thorium cycle). The last part of the article presents a preliminary safety analysis of the MSFR. (A.C.)

Oklo natural reactor

International Nuclear Information System (INIS)

Fujii, Isao

1985-01-01

In 1954, Professor Kazuo, Kuroda of Arkansas University in USA published the possibility that spontaneously generated natural nuclear reactors existed in prehistoric age. In 1972, 18 years after that, Commissariat a l'Energie Atomique published that in the Oklo uranium deposit in Gabon, Africa, a natural nuclear reactor was found. This fact was immediately informed to the whole world, but in Japan, its details have not necessarily been well known. The chance of investigating into this fact and visiting the Oklo deposit by the favor of COMUF, the owner of the Oklo deposit, was given, therefore, the state of the natural reactors, which has been known so far, is reported. At present, 12 natural reactors have been found in the vicinity of the Oklo deposit. The natural reactors were generated spontaneously in uranium deposits about 1.7 billion years ago when the isotopic abundance of U-235 was 3 %, and the chain reaction started naturally. When the concentration of U-235 lowered, the reaction stopped naturally. The abnormality in the U-235 abundance in natural uranium was found, and the cause was pursued. The evidence of the existence of natural reactors was shown. (Kako, I.)

Preparation of cellulose nanocrystals from asparagus (Asparagus officinalis L.) and their applications to palm oil/water Pickering emulsion.

Science.gov (United States)

Wang, Wenhang; Du, Guanhua; Li, Cong; Zhang, Hongjie; Long, Yunduo; Ni, Yonghao

2016-10-20

Nano cellulosic materials as promising emulsion stabilizers have attracted great interest in food industry. In this paper, five different sized cellulose nanocrystals (CNC) samples were prepared from stem of Asparagus officinalis L. using the same sulfuric acid hydrolysis conditions but different times (1.5, 2, 2.5, 3.0, and 3.5h). The sizes of these CNC ranged from 178.2 to 261.8nm, with their crystallinity of 72.4-77.2%. The CNC aqueous dispersions showed a typical shear thinning behavior. In a palm oil/water (30/70, v/v) model solution, stable Pickering emulsions were formed with the addition of CNC, and their sizes are in the range of 1-10μm based on the optical and confocal laser scanning microscopy (CLSM) observation. The CNC sample prepared at 3h hydrolysis time, showed a relative efficient emulsion capacity for palm oil droplets, among these CNCs. Other parameters including the CNC, salt, and casein concentrations on the emulsion stability were studied. Copyright © 2016 Elsevier Ltd. All rights reserved.

Anemia em catadores de material reciclável que utilizam carrinho de propulsão humana no município de Santos Anemia in recyclable waste pickers using human driven pushcarts in the city of Santos, southeastern Brazil

Directory of Open Access Journals (Sweden)

Mauro Abrahão Rozman

2010-06-01

Full Text Available OBJETIVO: Estimar a prevalência de anemia e analisar os fatores de risco a ela associados nos catadores de material reciclável que utilizam carrinho de propulsão humana do município de Santos - São Paulo. MÉTODO: Estudo transversal com 253 catadores foi realizado em julho de 2005. A coleta de informações foi feita por meio de questionário com informações sobre características individuais, ocupacionais e dietéticas. Foi realizada avaliação antropométrica e coletado sangue venoso para hemograma completo e sorologias de HIV, HCV, HBV e sífilis. A análise estatística foi feita por análise uni e multivariada (regressão logística, relacionando a anemia aos fatores de risco. RESULTADOS: A prevalência de anemia foi de 38,3%. As variáveis que mostraram associação independente com anemia no modelo multivariado foram: sexo (OR 4,11; IC95%: 1,56-10,87, infecção pelo HIV (OR 9,23; IC95%: 2,93-29,1, IMC (OR 0,21; IC95%: 0,07-0,64, anos de trabalho como catador (OR 4,54; IC95%: 1,29-16,0, consumo de leite (OR 0,36; IC95%: 0,16-0,81 e de proteína animal (OR 0,39; IC95%: 0,15-0,97. CONCLUSÃO: A prevalência de anemia entre catadores de material reciclável é elevada mesmo após a obrigatoriedade de adição de ferro nas farinhas de trigo e milho. Os catadores são excluídos das ações de proteção à saúde do trabalhador, previstas na legislação. Ações de saúde dirigidas a essa categoria profissional devem ser implementadas, garantindo a acessibilidade aos serviços de saúde.OBJECTIVE: To assess the prevalence of anemia and describe associated risk factors in recyclable waste pickers using human-driven pushcarts in the city of Santos. METHODS: A cross-sectional study including 253 recyclable waste pickers was conducted in the city of Santos, southeastern Brazil, in July 2005. A questionnaire was used to collect information about individual, occupational, and dietary factors. All subjects underwent an anthropometric

Enhancing mass transfer and ethanol production in syngas fermentation of Clostridium carboxidivorans P7 through a monolithic biofilm reactor

International Nuclear Information System (INIS)

Shen, Yanwen; Brown, Robert; Wen, Zhiyou

2014-01-01

Highlights: • Syngas fermentation process is limited by gas-to-liquid mass transfer. • A novel monolithic biofilm reactor (MBR) for efficient mass transfer was developed. • MBR with slug flow resulted in higher k L a than bubble column reactor (BCR). • MBR enhanced ethanol productivity by 53% compared to BCR. • MBR was demonstrated as a promising reactor configuration for syngas fermentation. - Abstract: Syngas fermentation is a promising process for producing fuels and chemicals from lignocellulosic biomass. Currently syngas fermentation faces several engineering challenges, with gas-to-liquid mass transfer limitation representing the major bottleneck. The aim of this work is to evaluate the performance of a monolithic biofilm reactor (MBR) as a novel reactor configuration for syngas fermentation. The volumetric mass transfer coefficient (k L a) of the MBR was evaluated in abiotic conditions within a wide range of gas flow rates (i.e., gas velocity in monolithic channels) and liquid flow rates (i.e., liquid velocity in the channels). The k L a values of the MBR were higher than those of a controlled bubble column reactor (BCR) in certain conditions, due to the slug flow pattern in the monolithic channels. A continuous syngas fermentation using Clostridium carboxidivorans P7 was conducted in the MBR system under varying operational conditions, with the variables including syngas flow rate, liquid recirculation between the monolithic column and reservoir, and dilution rate. It was found that the syngas fermentation performance – measured by such parameters as syngas utilization efficiency, ethanol concentration and productivity, and ratio of ethanol to acetic acid – depended not only on the mass transfer efficiency but also on the biofouling or abrading of the biofilm attached on the monolithic channel wall. At a condition of 300 mL/min of syngas flow rate, 500 mL/min of liquid flow rate, and 0.48 day −1 of dilution rate, the MBR produced much higher

Annual report, 1978

International Nuclear Information System (INIS)

1979-04-01

In 1978, for the first time, nuclear generators supplied more electricity than coal-fired units: 30 percent of the total compared with 28 percent for coal. Energy demand in Ontario was up by 2.7 percent. No new commitments for generating stations were made, and work on committed stations was to be slowed until the generation expansion program had been fully reviewed. Atomic Energy of Canada Ltd. and Ontario Hydro have agreed to develop a nuclear wastes demonstration facility. The Select Committee of the Ontario Legislature on Ontario Hydro affairs investigated the costs of the Bruce heavy water plants and found no evidence of mismanagement. The Royal Commission on Electric Power Planning issued an interim report on nuclear power which recognized the need for and safety of the CANDU system. Reactors at Pickering and Bruce achieved an overall capacity factor of 81 percent. The third Bruce A unit was started up. Work on Bruce B and Pickering B was well underway. Bruce Heavy Water Plant B was virtually complete, but work was stopped on the second half of the Bruce D plant. Plans for the first half of Bruce D will be reviewed. Site preparation and excavation continued for the Darlington generating station. (LL)

Some novel on-power refuelling features of CANDU stations

International Nuclear Information System (INIS)

Erwin, D.; Pendlebury, B.; Watson, J.F.; Welch, A.C.

1976-01-01

Part A of the paper describes the reasons for, and advantages resulting from, the use of flow assisted refuelling in the CANDU type nuclear reactors at the Pickering Generating Station. A separate fuel handling system is used for each reactor unit, as distinct from the system employed at the Bruce Generating station, where the fuel handling system is shared among several units. Part B of the paper describes some of the advantages of the shared concept with particular emphasis on the availability of the fuel handling system. (author)

Molecularly imprinted polymer microspheres prepared by Pickering emulsion polymerization for selective solid-phase extraction of eight bisphenols from human urine samples.

Science.gov (United States)

Yang, Jiajia; Li, Yun; Wang, Jincheng; Sun, Xiaoli; Cao, Rong; Sun, Hao; Huang, Chaonan; Chen, Jiping

2015-05-04

The bisphenol A (BPA) imprinted polymer microspheres were prepared by simple Pickering emulsion polymerization. Compared to traditional bulk polymerization, both high yields of polymer and good control of particle sizes were achieved. The characterization results of scanning electron microscopy and nitrogen adsorption-desorption measurements showed that the obtained molecularly imprinted polymer microsphere (MIPMS) particles possessed regular spherical shape, narrow diameter distribution (30-60 μm), a specific surface area (S(BET)) of 281.26 m(2) g(-1) and a total pore volume (V(t)) of 0.459 cm(3) g(-1). Good specific adsorption capacity for BPA was obtained in the sorption experiment and good class selectivity for BPA and its seven structural analogs (bisphenol F, bisphenol B, bisphenol E, bisphenol AF, bisphenol S, bisphenol AP and bisphenol Z) was demonstrated by the chromatographic evaluation experiment. The MIPMS as solid-phase extraction (SPE) packing material was then evaluated for extraction and clean-up of these bisphenols (BPs) from human urine samples. An accurate and sensitive analytical method based on the MIPMS-SPE coupled with HPLC-DAD has been successfully established for simultaneous determination of eight BPs from human urine samples with detection limits of 1.2-2.2 ng mL(-1). The recoveries of BPs for urine samples at two spiking levels (100 and 500 ng mL(-1) for each BP) were in the range of 81.3-106.7% with RSD values below 8.3%. Copyright © 2015 Elsevier B.V. All rights reserved.

Mirror reactor studies

International Nuclear Information System (INIS)

Moir, R.W.; Barr, W.L.; Bender, D.J.

1977-01-01

Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

Calculation of the transmutation rates of Tc-99, I-129 and Cs-135 in the High Flux Reactor, in the Phenix Reactor and in a light water reactor

International Nuclear Information System (INIS)

Bultman, J.

1992-04-01

Transmutation of long-lived fission products is of interest for the reduction of the possible dose to the population resulting from long-term leakage of nuclear waste from waste disposals. Three isotopes are of special interest: Tc-99, I-129 and Cs-135. Therefore, experiments on transmutation of these isotopes in nuclear reactors are planned. In the present study, the possible transmutation rates and mass reductions are determined for experiments in High Flux Reactor (HFR) located in Petten (Netherlands) and in Phenix (France). Also, rates were determined for a standard Light Water Reactor (LWR). The transmutation rates of the 3 fission products will be much higher in HFR than in Phenix reactor, as both total flux and effective cross sections are higher. For thick targets the effective half lives are approximately 3, 2 and 7 years for Tc-99, I-129 and Cs-135 irradiation respectively in HFR and 22, 16 and 40 years for Tc-99, I-129 and Cs-135 irradiation in Phenix reactor. The transmutation rates in LWR are low. Only the relatively large power of LWR guarantees a large total mass reduction. Especially transmutation of Cs-135 will be very difficult in Phenix and LWR, clearly shown by the very long effective half lives of 40 and 100 years, respectively. (author). 7 refs.; 5 figs.; 7 tabs

Structure and thermal properties of as-fabricated U-7Mo/Mg and U-10Mo/Mg low-enriched uranium research reactor fuels

Energy Technology Data Exchange (ETDEWEB)

Kulakov, Mykola, E-mail: mykola.kulakov@cnl.ca [Fuel Development Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Saoudi, Mouna [Fuel Development Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Piro, Markus H.A. [Fuel and Fuel Channel Safety Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Donaberger, Ronald L. [Canadian Neutron Beam Centre, Chalk River, ON K0J 1J0 Canada (Canada)

2017-02-15

Aluminum-clad U-7Mo/Mg and U-10Mo/Mg pin-type mini-elements (with a core uranium loading of 4.5 gU/cm{sup 3}) have been fabricated at the Canadian Nuclear Laboratories for experimental tests and ultimately for use in research and test reactors. In this study, the microstructure and phase composition of unirradiated U-7Mo/Mg and U-10Mo/Mg fuel cores were analyzed using optical and scanning electron microscopy, and neutron powder diffraction. Thermal properties were characterized using a combination of experimental measurements and thermodynamic calculations. The thermal diffusivity was measured using the laser flash method. The temperature-dependent specific heat capacities were calculated based on the linear rule of mixture using the weight fraction of different crystalline phases and their specific heat capacity values taken from the literature. The thermal conductivity was then calculated using the measured thermal diffusivity, the measured density and the calculated specific heat capacity. The resulting thermal conductivity is practically identical for both types of fuel. The in-reactor temperatures were predicted using conjugate heat transfer simulations. - Highlights: • Neutron diffraction analysis shows that most of the γ-U(Mo) phase was retained in as-fabricated U-7Mo/Mg and U-10Mo/Mg fuel cores. • The experimental thermal conductivity of both types of fuel is practically identical. • Based on conjugate heat transfer simulations, under normal operating conditions, the in-reactor fuel centreline temperature is about 510 K.

Progress report concerning safety research for nuclear reactor facilities

International Nuclear Information System (INIS)

1978-01-01

Examination and evaluation of safety research results for nuclear reactor facilities have been performed, as more than a year has elapsed since the plan had been initiated in April, 1976, by the special sub-committee for the safety of nuclear reactor facilities. The research is carried out by being divided roughly into 7 items, and seems to be steadily proceeding, though it does not yet reach the target. The above 7 items include researches for (1) criticality accident, (2) loss of coolant accident, (3) safety for light water reactor fuel, (4) construction safety for reactor facilities, (5) reduction of release of radioactive material, (6) safety evaluation based on the probability theory for reactor facilities, and (7) aseismatic measures for reactor facilities. With discussions on the progress and the results of the research this time, research on the behaviour on fuel in abnormal transients including in-core and out-core experiments has been added to the third item, deleting the power-cooling mismatch experiment in Nuclear Safety Research Reactor of JAERI. Also it has been decided to add two research to the seventh item, namely measured data collection, classification and analysis, and probability assessment of failures due to an earthquake. For these 7 items, the report describes the concrete contents of research to be performed in fiscal years of 1977 and 1978, by discussing on most rational and suitable contents conceivable at present. (Wakatsuki, Y.)

1: the atom. 2: radioactivity. 3: man and radiations. 4: the energy. 5: nuclear energy: fusion and fission. 6: the operation of a nuclear reactor. 7: the nuclear fuel cycle

International Nuclear Information System (INIS)

2002-01-01

This series of 7 digest booklets present the bases of the nuclear physics and of the nuclear energy: 1 - the atom (structure of matter, chemical elements and isotopes, the four fundamental interactions, nuclear physics); 2 - radioactivity (definition, origins of radioelements, applications of radioactivity); 3 - man and radiations (radiations diversity, biological effects, radioprotection, examples of radiation applications); 4 - energy (energy states, different forms of energy, characteristics); 5 - nuclear energy: fusion and fission (nuclear energy release, thermonuclear fusion, nuclear fission and chain reaction); 6 - operation of a nuclear reactor (nuclear fission, reactor components, reactor types); 7 - nuclear fuel cycle (nuclear fuel preparation, fuel consumption, reprocessing, wastes management). (J.S.)

Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

Energy Technology Data Exchange (ETDEWEB)

Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science in Radiation and Dynamics Extremes

2016-09-26

In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

Hétérogenéité quantitative et qualitative de la matière organique dans les argiles du Kimmeridgien du val de Pickering (Yorkshire, UK). Cadre sédimentologique et stratigraphique Quantitative and Qualitative Heterogeneity of the Organic Matter Within the Kimmeridge Clay of the Vale of Pickering (Yorkshire, Uk). Sedimentological and Stratigraphical Framework

OpenAIRE

Penn I. E.; Herbin J. P.; Muller C.; Geyssant J. R.; Melieres F.

2006-01-01

Les argiles du Kimméridgien contribuent à l'alimentation de nombreux gisements en mer du Nord. L'étude de quatre puits localisés sur une coupe Est-Ouest dans le Val de Pickering (Bassin de Cleveland) permet d'illustrer dans cette formation l'hétérogénéité de distribution verticale et latérale de la matière organique en terme de quantité et de qualité. La coupe étudiée (35 km de long, 200 m de haut) représente environ 6,5 millions d'années d'histoire géologique. Les analyses minéralogiques mon...

Fusion reactors as a future energy source

International Nuclear Information System (INIS)

Seifritz, W.

A detailed update of fusion research concepts is given. Discussions are given for the following areas: (1) the magnetic confinement principle, (2) UWMAK I: conceptual design for a fusion reactor, (3) the inertial confinement principle, (4) the laser fusion power plant, (5) electron-induced fusion, (6) the long-term development potential of fusion reactors, (7) the symbiosis between fusion and fission reactors, (8) fuel supply for fusion reactors, (9) safety and environmental impact, and (10) accidents, and (11) waste removal and storage

RA reactor exploitation, task 3.08/01

International Nuclear Information System (INIS)

Zecevic, V.

1963-01-01

During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report

Simulation development for TRIGA reactor

International Nuclear Information System (INIS)

Handoyo, D.

1997-01-01

A simulator of the dynamic of TRIGA reactor has been made. this simulator is meant to study the reactor kinetic behavior and for operator training to more assure the safety and the reliability of the real operation of TRIGA reactor. the simulator consists of PC (Personal Computer) for processing the calculation of reactivity, neutron flux, period, ect and control panel for regulating the input data such as the change of power range, control rod position as well as cooling flow rate. the result will be displayed on screen monitor of personal computer as given in the real control room of TRIGA reactor. the output of simulator will be verified by comparing with measurement result in the real TRIGA MARK II reactor of Musashi institute of technology. for the change of reactivity of 0.3, 0.5 and 0.7 the reactor power and fuel temperature between the simulator and measurements are comparable

Fast Reactor Physics. Vol. II. Proceedings of a Symposium on Fast Reactor Physics and Related Safety Problems

International Nuclear Information System (INIS)

1968-01-01

Proceedings of a Symposium organized by the IAEA and held in Karlsruhe, 30 October - 3 November 1967. The meeting was attended by 183 scientists from 23 countries and three international organizations. Contents: (Vol.1) Review of national programmes (5 papers); Nuclear data for fast reactors (12 papers); Experimental methods (3 papers); Zoned systems (7 papers); Kinetics (7 papers). (Vol.11) Fast critical experiments (8 papers); Heterogeneity in fast critical experiments (5 papers); Fast power reactors (13 papers); Fast pulsed reactors (3 papers); Panel discussion. Each paper is in its original language (50 English, 11 French and 3 Russian) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

Graphite-water steam-generating reactor in the USSR

Energy Technology Data Exchange (ETDEWEB)

Dollezhal, N A [AN SSSR, Moscow

1981-10-01

One of the types of power reactor used in the USSR is the graphite-water steam-generating reactor RBMK. This produces saturated steam at a pressure of 7MPa. Reactors giving 1GWe each have been installed at the Leningrad, Kursk, Chernobyl and other power stations. Further stations using reactors of this type are being built. A description is given of the fuel element design, and of the layout of the plant. The main characteristics of RBMK reactors using fuel of rated and higher enrichment are listed.

Dose rate in the reactor room and environment during maintenance in fusion reactors

International Nuclear Information System (INIS)

Maki, Koichi; Satoh, Satoshi; Takatsu, Hideyuki; Seki, Yasushi

1995-01-01

According to the International Thermonuclear Experimental Reactor (ITER) conceptual design activity, after reactor shutdown, damaged segments are pulled up from the reactor and hung from the reactor room ceiling by a remote handling device. The dose rate in the reactor room and the environment is estimated for this situation, and the following results are obtained. First, the dose rate in the room is > 10 8 μSv/h. Since this dose rate is 10 7 times greater than the biological radiation shielding design limit of 25 μSv/h, workers cannot enter the room. Second, lenses and optical fiber composed of glass that is radiation resistant up to 10 6 Gy would be damaged after <100 h near the segment, and devices using semiconductors could not work after several hours or so in the aforementioned dose-rate conditions. Third, during suspension of one blanket segment from the ceiling, the dose rate in the site boundary can be reduced by one order by a 23-cm-thicker reactor building roof. To reduce dose rate in public exposure to a value that is less than one-tenth of the public exposure radiation shielding design limit of 100 μSv/yr, the distance of the site boundary from the reactor must be greater than 200 m for a reactor building with a 160-cm-thick concrete roof. 9 refs., 6 figs., 2 tabs

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

2011-08-15

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

Expected characteristics of future reactors for human beings

International Nuclear Information System (INIS)

Taketani, Kiyoaki

1992-01-01

Based on four reactor safety components (namely: a) God-given safety, b) Equipment safety, c) Quick-response safety, d) Containing safety), categorical assessment is made of various nuclear reactor concepts ranging from present existing reactors to future reactors based on innovative reactor design. In pursuit of nuclear reactor safety, ultimate characteristics of the ideal nuclear reactor are expected to coincide with those of an inherently safe reactor. A definition of 'inherently safe' has already been proposed by a committee in Japan. As a realistic and existable reactor, which is as close to the ideal reactor, a future reactor which is almost the same as a global reactor, is proposed. This global reactor must be constructable anywhere on earth and must permit easy operation and maintenance by anyone. It is also discussed to identify what behavior is expected of the global reactor under various conditions. At the same time, this future reactor which includes the global reactor, should solve a) the nuclear fuel resource issue, b) efficient utilization of fission energy and c) environmental issues as the greenhouse effect. (author). 7 refs., 2 figs

RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.; Miokovic, J.

1982-12-01

Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

Practical course on reactor instrumentation

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2004-06-01

This course is based on the description of the instrumentation of the TRIGA-reactor Vienna, which is used for training research and isotope production. It comprises the following chapters: 1. instrumentation, 2. calibration of the nuclear channels, 3. rod drop time of the control rods, 4. neutron flux density measurements using compensated ionization, 5. neutron flux density measurement with fission chambers (FC), 6. neutron flux density measurement with self-powered neutron detectors (SPND), 7. pressurized water reactor simulator, 8. verification of the radiation level during reactor operation. There is one appendix about neutron-sensitive thermocouples. (nevyjel)

Highlights of the metallurgical behaviour of CANDU pressure tubes

International Nuclear Information System (INIS)

Price, E.G.

1984-10-01

This paper is an overview of the service induced metallurgical changes that take place in Zircaloy-2 and Zr-2.5 wt. percent Nb pressure tubes in CANDU reactors. It incorporates the findings of an evaluation program, that followed a significant pressure tube failure at Ontario Hydro's Pickering Nuclear Generating Station, and also provides valid reasons for continued confidence in the current CANDU design

Fabrication and evaluation of chitosan/NaYF{sub 4}:Yb{sup 3+}/Tm{sup 3+} upconversion nanoparticles composite beads based on the gelling of Pickering emulsion droplets

Energy Technology Data Exchange (ETDEWEB)

Yan, Huiqiong; Chen, Xiuqiong; Shi, Jia; Shi, Zaifeng; Sun, Wei; Lin, Qiang, E-mail: linqianggroup@163.com; Wang, Xianghui; Dai, Zihao

2017-02-01

The rare earth ion doped upconversion nanoparticles (UCNPs) synthesized by hydrophobic organic ligands possess poor solubility and low fluorescence quantum yield in aqueous media. To conquer this issue, NaYF{sub 4}:Yb{sup 3+}/Tm{sup 3+} UCNPs, synthesized by a hydrothermal method, were coated with F127 and then assembled with chitosan to fabricate the chitosan/NaYF{sub 4}:Yb{sup 3+}/Tm{sup 3+} composite beads (CS/NaYF{sub 4}:Yb{sup 3+}/Tm{sup 3+} CBs) by Pickering emulsion system. The characterization results revealed that the as-synthesized NaYF{sub 4}:Yb{sup 3+}/Tm{sup 3+} UCNPs with an average size of 20 nm exhibited spherical morphology, high crystallinity and characteristic emission upconversion fluorescence with an overall blue color output. The NaYF{sub 4}:Yb{sup 3+}/Tm{sup 3+} UCNPs were successfully conjugated on the surface of chitosan beads by the gelling of emulsion droplets. The resultant CS/NaYF{sub 4}:Yb{sup 3+}/Tm{sup 3+} CBs showed good upconversion luminescent property, drug-loading capacity, release performance and excellent biocompatibility, exhibiting great potentials in targeted drug delivery and tissue engineering with potential tracking capability and lasting release performance. - Highlights: • NaYF{sub 4}:Yb{sup 3+}/Tm{sup 3+} UCNPs were coated by F127 to improve aqueous dispersibility. • NaYF{sub 4}:Yb{sup 3+}/Tm{sup 3+} UCNPs were assembled with chitosan to fabricate the composite beads (CMs). • Pickering emulsions stabilized by UCNPs exhibited uniform and satisfactory emulsion droplets. • The CMs prepared by the gelling of emulsion droplet preserved upconversion luminescent property. • The resultant CMs showed good drug-loading capacity, release performance and biocompatibility.

Upper limit to the deuterium abundance and a measurement of the pickering-β line in the low excitation planetary IC 418

International Nuclear Information System (INIS)

Le Vaux, H.A.

1977-01-01

The problem of detecting a weak spectrum line, deuterium alpha, very near in wavelength to H/sub alpha/, assumed to be thousands of times as strong, is discussed from the point of view of optimizing the signal-to-noise ratio. A spectrometer consisting of three pressure scanning Fabry-Perot etalons with low reflectivity coatings was found to be the best instrument for this experiment. While no feature attributable to deuterium was detected in the planetary nebula IC 418, an upper limit relative to hydrogen of 3.4 x 10 -5 was established at the 95% confidence level. The significance of this result is discussed in light of the role played by deuterium in cosmology. The intensity ratio of the Pickering beta line (n'' = 6, n' = 4 transition of ionized helium) relative to H/sub alpha/ was measured to be 6.5 x 10 -5 . Observations of the nebular continuum made at five wavelengths in the red and near infrared are summarized and compared with predicted intensities

Digital control system of advanced reactor

International Nuclear Information System (INIS)

Peng Huaqing; Zhang Rui; Liu Lixin

2001-01-01

This article produced the Digital Control System For Advanced Reactor made by NPIC. This system uses Siemens SIMATIC PCS 7 process control system and includes five control system: reactor power control system, pressurizer level control system, pressurizer pressure control system, steam generator water level control system and dump control system. This system uses three automatic station to realize the function of five control system. Because the safety requisition of reactor is very strict, the system is redundant. The system configuration uses CFC and SCL. the human-machine interface is configured by Wincc. Finally the system passed the test of simulation by using RETRAN 02 to simulate the control object. The research solved the key technology of digital control system of reactor and will be very helpful for the nationalization of digital reactor control system

Moderator heat recovery of CANDU reactors

International Nuclear Information System (INIS)

Fath, H.E.S.; Ahmed, S.T.

1986-01-01

A moderator heat recovery scheme is proposed for CANDU reactors. The proposed circuit utilizes all the moderator heat to the first stages of the plant feedwater heating system. CANDU-600 reactors are considered with moderator heat load varying from 120 to 160 MWsub(th), and moderator outlet temperature (from calandria) varying from 80 to 100 0 C. The steam saved from the turbine extraction system was found to produce an additional electric power ranging from 5 to 11 MW. This additional power represents a 0.7-1.7% increase in the plant electric output power and a 0.2-0.7% increase in the plant thermal efficiency. The outstanding features and advantages of the proposed scheme are presented. (author)

Modeling and performance of the MHTGR [Modular High-Temperature Gas-Cooled Reactor] reactor cavity cooling system

International Nuclear Information System (INIS)

Conklin, J.C.

1990-04-01

The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab

Introduction to reactor internal materials for pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-06-01

This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

Introduction to reactor internal materials for pressurized water reactor

International Nuclear Information System (INIS)

Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun

1994-06-01

This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,

Hanford B Reactor Building Hazard Assessment Report

International Nuclear Information System (INIS)

Griffin, P. W.

1999-01-01

The 105-B Reactor (hereinafter referred to as B Reactor) is located in the 100 Area of the Hanford Site near Richland, Washington. The B Reactor is one of nine plutonium production reactors that were constructed in the 1940s during the Cold War Era. Construction of the B Reactor began June 7, 1943, and operation began on September 26, 1944. The Environmental Restoration Contractor was requested by RL to provide an assessment/characterization of the B Reactor building to determine and document the hazards that are present and could pose a threat to the environment and/or to individuals touring the building. This report documents the potential hazards, determines the feasibility of mitigating the hazards, and makes recommendations regarding areas where public tour access should not be permitted

Recent fuel handling experience in Canada

International Nuclear Information System (INIS)

Welch, A.C.

1991-01-01

For many years, good operation of the fuel handling system at Ontario Hydro's nuclear stations has been taken for granted with the unavailability of the station arising from fuel handling system-related problems usually contributing less than one percent of the total unavailability of the stations. While the situation at the newer Hydro stations continues generally to be good (with the specific exception of some units at Pickering B) some specific and some general problems have caused significant loss of availability at the older plants (Pickering A and Bruce A). Generally the experience at the 600 MWe units in Canada has also continued to be good with Point Lepreau leading the world in availability. As a result of working to correct identified deficiencies, there were some changes for the better as some items of equipment that were a chronic source of trouble were replaced with improved components. In addition, the fuel handling system has been used three times as a delivery system for large-scale non destructive examination of the pressure tubes, twice at Bruce and once at Pickering and performing these inspections this way has saved many days of reactor downtime. Under COG there are several programs to develop improved versions of some of the main assemblies of the fuelling machine head. This paper will generally cover the events relating to Pickering in more detail but will describe the problems with the Bruce Fuelling Machine Bridges since the 600 MW 1P stations have a bridge drive arrangement that is somewhat similar to Bruce

Results and recommendations from the reactor chemistry and corrosion tasks of the reactor materials program

International Nuclear Information System (INIS)

Baumann, E.W.; Ondrejcin, R.S.

1990-11-01

Within the general context of extended service life, the Reactor Materials Program was initiated in 1984. This comprehensive program addressed material performance in SRS reactor tanks and the primary coolant or Process Water System (PWS) piping. Three of the eleven tasks concerned moderator quality and corrosion mitigation. Definition and control of the stainless steel aqueous environment is a key factor in corrosion mitigation. The Reactor Materials Program systematically investigated the SRS environment and its effect on crack initiation and propagation in stainless steel, with the objective of improving this environment. The purpose of this report is to summarize the contributions of Tasks 6, 7 and 10 of the Reactor Materials Program to the understanding and control of moderator quality and its relationship to mitigation of stress corrosion cracking

Selective separation of lambdacyhalothrin by porous/magnetic molecularly imprinted polymers prepared by Pickering emulsion polymerization.

Science.gov (United States)

Hang, Hui; Li, Chunxiang; Pan, Jianming; Li, Linzi; Dai, Jiangdong; Dai, Xiaohui; Yu, Ping; Feng, Yonghai

2013-10-01

Porous/magnetic molecularly imprinted polymers (PM-MIPs) were prepared by Pickering emulsion polymerization. The reaction was carried out in an oil/water emulsion using magnetic halloysite nanotubes as the stabilizer instead of a toxic surfactant. In the oil phase, the imprinting process was conducted by radical polymerization of functional and cross-linked monomers, and porogen chloroform generated steam under the high reaction temperature, which resulted in some pores decorated with easily accessible molecular binding sites within the as-made PM-MIPs. The characterization demonstrated that the PM-MIPs were porous and magnetic inorganic-polymer composite microparticles with magnetic sensitivity (M(s) = 0.7448 emu/g), thermal stability (below 473 K) and magnetic stability (over the pH range of 2.0-8.0). The PM-MIPs were used as a sorbent for the selective binding of lambdacyhalothrin (LC) and rapidly separated under an external magnetic field. The Freundlich isotherm model gave a good fit to the experimental data. The adsorption kinetics of the PM-MIPs was well described by pseudo-second-order kinetics, indicating that the chemical process could be the rate-limiting step in the adsorption of LC. The selective recognition experiments exhibited the outstanding selective adsorption effect of the PM-MIPs for target LC. Moreover, the PM-MIPs regeneration without significant loss in adsorption capacity was demonstrated by at least four repeated cycles. © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

University Reactor Instrumentation grant program. Final report, September 7, 1990--August 31, 1995

International Nuclear Information System (INIS)

Talnagi, J.W.

1998-01-01

The Ohio State University Nuclear Reactor Laboratory (OSU NRL) participated in the Department of Energy (DOE) grant program commonly denoted as the University Reactor Instrumentation (URI) program from the period September 1990 through August 1995, after which funding was terminated on a programmatic basis by DOE. This program provided funding support for acquisition of capital equipment targeted for facility upgrades and improvements, including modernizing reactor systems and instrumentation, improvements in research and instructional capabilities, and infrastructure enhancements. The staff of the OSU NRL submitted five grant applications during this period, all of which were funded either partially or in their entirety. This report will provide an overview of the activities carried out under these grants and assess their impact on the OSU NRL facilities

Recovery of tritium from CANDU reactors, its storage and monitoring of its migration in the environment

International Nuclear Information System (INIS)

Holtslander, W.J.; Osborne, R.V.

1979-07-01

Tritium is produced in CANDU heavy water reactors mainly by neutron activation of deuterium. The typical production rate is 2.4 kCi per megawatt-year (89 TBq. per megawatt-year. In Pickering Generating Station the average concentration of tritium in the moderators has reached 16 Ci.kg -1 (0.6 TBq.kg -1 ) and in coolants, 0.5 Ci.kg -1 (0.02 TBq.kg -1 ). Concentrations will continue to increase towards an equilibrium determined by the production rate, the tritium decay rate and heavy water replacement. Tritium removal methods that are being considered for a pilot plant design are catalytic exchange of DTO with D 2 and electrolysis of D 2 O/DTO to provide feed for cryogenic distillation of D 2 /DT/T 2 . Storage methods for the removed tritium - as elemental gas, as metal hydrides and in cements - are also being investigated. Transport of tritiated wastes should not be a particularly difficult problem in light of extensive experience in transporting tritiated heavy water. Methods for determining the presence of tritium in the environment of any tritium handling facility are well established and have the capability of measuring concentrations of tritium down to current ambient values. (author)

The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

Energy Technology Data Exchange (ETDEWEB)

Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Nuclear Reactor Lab.; Wilson, Erik [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-12-01

The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.

FISA 2009 - 7th European Commission conference on EURATOM research and training in reactor systems. Conference proceedings

International Nuclear Information System (INIS)

Goethem, G. van; Manolatos, P.; Hugon, M.; Bhatnagar, V.; Deffrennes, M.; Webster, S.

2010-01-01

The main achievements of the first series of projects under EURATOM FP-7 for nuclear research and training activities (2007 to 2011) were discussed. Approximately 500 participants were registered at FISA 2009 and at the 7 post-conference workshops, representing a wide audience of nuclear scientists and decision makers coming from 32 countries worldwide. The focus of the conference was on scientific and technological research in the following areas: nuclear plant life management for existing reactors (Generation II), severe accident management (Generation III), assessment of future nuclear fission systems (Generation IV), partitioning and transmutation systems (innovative fuels), access to large research infrastructures, and nuclear education and training. Special attention was devoted to the societal and industrial goals of GIF: sustainability, industrial competitiveness, safety and reliability, proliferation resistance. (orig.)

MLR reactor

International Nuclear Information System (INIS)

Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

1998-01-01

The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

The SPHINX reactor for engineering tests

International Nuclear Information System (INIS)

Adamov, E.O.; Artamkin, K.N.; Bovin, A.P.; Bulkin, Y.M.; Kartashev, E.F.; Korneev, A.A.; Stenbok, I.A.; Terekhov, A.S.; Khmel'Shehikov, V.V.; Cherkashov, Y.M.

1990-01-01

A research reactor known as SPHINX is under development in the USSR. The reactor will be used mainly to carry out tests on mock-up power reactor fuel assemblies under close-to-normal parameters in experimental loop channels installed in the core and reflector of the reactor, as well as to test samples of structural materials in ampoule and loop channels. The SPHINX reactor is a channel-type reactor with light-water coolant and moderator. Maximum achievable neutron flux density in the experimental channels (cell composition 50% Fe, 50% H 2 O) is 1.1 X 10 15 neutrons/cm 2 · s for fast neutrons (E > 0.1 MeV) and 1.7 X 10 15 for thermal neutrons at a reactor power of 200 MW. The design concepts used represent a further development of the technical features which have met with approval in the MR and MIR channel-type engineering test reactors currently in use in the USSR. The 'in-pond channel' construction makes the facility flexible and eases the carrying out of experimental work while keeping discharges of radioactivity into the environment to a low level. The reactor and all associated buildings and constructions conform to modern radiation safety and environmental protection requirements

Submersion-Subcritical Safe Space (S4) reactor

International Nuclear Information System (INIS)

King, Jeffrey C.; El-Genk, Mohamed S.

2006-01-01

The Submersion-Subcritical Safe Space (S 4 ) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

Very-high-temperature reactors for future use

International Nuclear Information System (INIS)

Kasten, P.R.

1988-08-01

Very-high-temperature reactors (VHTRs) show promise for economic generation of electricity and of high-temperature process heat. The key is the development of high-temperature materials which permit gas turbine VHTRs to generate electricity economically, at reactor coolant temperatures which can be used for fossil fuel conversion processes. 7 refs., 5 figs

RAPK-7. code for calculating mass transfer and corrosion products activation in the circulation loops of water-cooled reactors

International Nuclear Information System (INIS)

Mikhaylov, A.V.; Moryakov, A.V.; Nikitin, A.V.

2012-09-01

The RAPK-7 code was developed to simulate formation of non-irradiated and activated corrosion products, their transport and deposition on inner surfaces of primary components and in primary coolant of water-cooled reactors during their operation on power and after shutdown. The key feature of this code is its particular emphasis on the contamination of circulation loops by radioactive corrosion products of reactor which operates on variable modes. Such reactors typically are: research reactors and their experimental loops, naval nuclear power systems, etc. It's typical for such reactors to have repeated (over the campaign) and frequent variations in power (activating neutron fluxes), thermal-physical, hydrodynamic and other parameters of coolant, intensive water mass exchange between the circulation loop and the pressuriser, etc. The processes of mass-transfer are described by the RAPK-7 code with the use of models similar to those employed by the COTRAN and PACTOLE codes. The circulation circuit is broken down into computation areas. The user will then set the concentrations of water chemistry adjusting additives (alkali, boric acid, ammonia, hydrogen), as well as parameters in each area, such as wall temperature, coolant flow core temperature, pressure, flow rate, velocity, the radial component of coolant flowrate and activating neutron flux density. All the above parameters can be set as time-dependent step functions (bar charts), with independent time steps for each of them. The number of computation areas, the number of time dependencies and the level of detail in their description are limited by computer capabilities only. A 'brake' mode with a single-step change of the required set of parameters is provided to allow for jump-type events, such as replacement of contaminated components with clean ones during core refueling or repairs, emergency injection of boric acid, water mass exchange between the circulation circuit and the pressuriser, etc

Behavior of low-burnup metallic fuels for the integral fast reactor at elevated temperatures in ex-reactor tests

International Nuclear Information System (INIS)

Tsai, Hanchung; Liu, Yung Y.; Wang, Da-Yung; Kramer, J.M.

1991-07-01

A series of ex-reactor heating tests on low burnup U-26wt.%Pu-10wt.%Zr metallic fuel for the PRISM reactor was conducted to evaluate fuel/cladding metallurgical interaction and its effect on cladding integrity at elevated temperatures. The reaction between the fuel and cladding caused liquid-phase formation and dissolution of the inner surface of the cladding. The rate of cladding penetration was below the existing design correlation, which provides a conservative margin to cladding failure. In a test which enveloped a wide range of postulated reactor transient events, a substantial temporal cladding integrity margin was demonstrated for an intact, whole fuel pin. The cause of the eventual pin breach was reaction-induced cladding thinning combined with fission-gas pressure loading. The behavior of the breached pin was benign. 7 refs., 7 figs., 1 tab

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Global risk of radioactive fallout after major nuclear reactor accidents

International Nuclear Information System (INIS)

Lelieveld, J.; Kunkel, D.; Lawrence, M.G.

2012-01-01

Major reactor accidents of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by ''rare''? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the cumulative, global risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents (the most severe ones on the International Nuclear Event Scale, INES 7), using particulate "1"3"7Cs and gaseous "1"3"1I as proxies for the fallout. Our results indicate that previously the occurrence of INES 7 major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a major reactor accident of any nuclear power plant worldwide, more than 90% of emitted "1"3"7Cs would be transported beyond 50 km and about 50% beyond 1000 km distance before being deposited. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of "1"3"7Cs and "1"3"1I are quite different, the radioactive contamination patterns over land and the human exposure due to deposition are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in West Europe and South Asia, where a major reactor accident can subject around 30 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

Assessment of torsatrons as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Painter, S.L.

1992-12-01

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R 0 = 6.6-8.8 m, on-axis magnetic field B 0 = 4.8-7.5 T, B max (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

Annual report on the state of RB reactor components and equipment, december 1998

International Nuclear Information System (INIS)

Milosevic, M.

1998-12-01

According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1998 the reactor lattice was changed 7 times, meaning that experiments were done with 7 configurations of the reactor core. Total reactor operation amounted to 177.5 Wh with 40 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment in 1998, reactor operation nd utilization data, plan for regular annual maintenance and refurbishment of reactor equipment and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

Effluent releases at the TRIGA reactor facility

Energy Technology Data Exchange (ETDEWEB)

Whittemore, W L [General Atomic Co., San Diego, CA (United States)

1974-07-01

The principal effluent from the operating TRIGA reactors in our facility is argon-41. As monitored by a recording gas and particulate stack monitor, the values shown in the table, the Mark III operating 24 hours per day for very long periods produced the largest amount of radioactive argon. The quantity of 23.7 Ci A-41 when diluted by the normal reactor room ventilation system corresponded to 1.45 x 10{sup -6} {mu}Ci/cc. As diluted in the roof stack stream and the reactor building wake, the concentration immediately outside the reactor building was 25% MPC for an unrestricted area. The continued dilution of this effluent resulted in a concentration of a few percent MPC at the site boundary (unrestricted area) 350 meters from the reactor. (author)

Target-fueled nuclear reactor for medical isotope production

Science.gov (United States)

Coats, Richard L.; Parma, Edward J.

2017-06-27

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

Energy Technology Data Exchange (ETDEWEB)

Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 ⁷LiF-BeF₂) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

International Nuclear Information System (INIS)

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.; Powers, Jeffrey J.

2016-01-01

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 "7LiF-BeF_2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

A two dimensional code (R,Z) for nuclear reactor analysis and its application to the UAR-RI reactor

International Nuclear Information System (INIS)

Bishay, S.T.; Mikhail, I.F.I.; Gaafar, M.A.; Michaiel, M.L.; Nassar, S.F.

1988-01-01

A detailed study is given of fuel consumption in completely reflected cylindrical reactors. A two group, two dimensional (r,z) code is developed to carry out the required procedure. The code is applied to the UAR-RI reactor and the results are found to be in complete agreement with the experimental observations and with the theoretical interpretations. 7 fig., 12 tab

YBa2Cu3O7-δ thin films deposited by MOCVD vertical reactor with a flow guide

International Nuclear Information System (INIS)

Sujiono, E.H.; Negeri Makassar; Sani, R.A.; Saragi, T.; Arifin, P.; Barmawi, M.

2001-01-01

The effect of a flow guide in a vertical MOCVD reactor on the deposition uniformity and growth rate of thin YBCO films has been studied. Without the flow guide the growth rates are low, have a poor uniformity and the film composition is not stoichiometric. The growth rate of the films grown using a reactor with the flow guide was approximately twice that without the flow guide. Using this flow guide the growth rates were 0.4-0.7 μm/h for growth temperatures varying between 600 and 750 C, and the crystalline quality as well as the surface morphology of YBCO films on MgO substrates is improved. For films grown at temperatures above 650 C the composition of Y:Ba:Cu is 1:2:3, as confirmed by EDAX spectra. Films deposited without and with the flow guide at 700 C have critical temperatures around 85 and 88 K, respectively. The reduction in ΔT c (T c,zero -T c,onset ) also shows an improvement of the superconducting properties of YBCO thin films deposited with a flow guide. (orig.)

Life extension, power upgrade, and return to service work for Pickering NGS and other PWR and CANDU plants

International Nuclear Information System (INIS)

Millman, J.; Idvorian, N.; Schneider, W.

2002-01-01

Work on life extension, power upgrade and return to service has been performed and is in progress for a number of PWR and CANDU plants. For PWR plants, power upgrade work has been done for the new replacement steam generators in several cases. This work consists of redoing the formal equipment qualification analysis and reports for the uprated operating conditions to support the application for license adjustment. Life extension assessments have been performed for several CANDU plants. These are highly detailed assessments in which the particular steam generator is reassessed part by part as to the ability of each to sustain full life operation and also extended life operation. Return to service work for Pickering NGSA specifically has included this type of assessment and also specific repair, cleaning and retrofit activities including secondary side inspection, waterlancing, divider plate repair, eddy current inspection, etc. Steam generator modifications and retrofit work have been performed in a number of cases. The paper discusses various life extension, power upgrade, equipment modification and return to service activities all of which are part of the renewed drive in the industry to realise the full potential of nuclear plants by getting more and better performance from the extended service of existing plants. (author)

Nuclear reactor refueling system

International Nuclear Information System (INIS)

Wade, E.E.

1978-01-01

A system for transferring fuel assemblies between a nuclear reactor core and a fuel storage area while the fuel assembies remain completely submerged in a continuous body of coolant is described. The system comprises an in-vessel fuel transfer machine located inside the reactor vessel and an ex-vessel fuel transfer machine located in a fuel storage tank. The in-vessel fuel transfer machine comprises two independently rotatable frames with a pivotable fuel transfer apparatus disposed on the lower rotatable frame. The ex-vessel fuel transfer machine comprises one frame with a pivotable fuel transfer apparatus disposed thereon. The pivotable apparatuses are capable of being aligned with each other to transfer a fuel assembly between the reactor vessel and fuel storage tank while the fuel assembly remains completely submerged in a continuous body of coolant. 9 claims, 7 figures

Interfacial rheology of model particles at liquid interfaces and its relation to (bicontinuous) Pickering emulsions

Science.gov (United States)

Thijssen, J. H. J.; Vermant, J.

2018-01-01

Interface-dominated materials are commonly encountered in both science and technology, and typical examples include foams and emulsions. Conventionally stabilised by surfactants, emulsions can also be stabilised by micron-sized particles. These so-called Pickering-Ramsden (PR) emulsions have received substantial interest, as they are model arrested systems, rather ubiquitous in industry and promising templates for advanced materials. The mechanical properties of the particle-laden liquid-liquid interface, probed via interfacial rheology, have been shown to play an important role in the formation and stability of PR emulsions. However, the morphological processes which control the formation of emulsions and foams in mixing devices, such as deformation, break-up, and coalescence, are complex and diverse, making it difficult to identify the precise role of the interfacial rheological properties. Interestingly, the role of interfacial rheology in the stability of bicontinuous PR emulsions (bijels) has been virtually unexplored, even though the phase separation process which leads to the formation of these systems is relatively simple and the interfacial deformation processes can be better conceptualised. Hence, the aims of this topical review are twofold. First, we review the existing literature on the interfacial rheology of particle-laden liquid interfaces in rheometrical flows, focussing mainly on model latex suspensions consisting of polystyrene particles carrying sulfate groups, which have been most extensively studied to date. The goal of this part of the review is to identify the generic features of the rheology of such systems. Secondly, we will discuss the relevance of these results to the formation and stability of PR emulsions and bijels.

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Neutron field control cybernetics model of RBMK reactor operator

International Nuclear Information System (INIS)

Polyakov, V.V.; Postnikov, V.V.; Sviridenkov, A.N.

1992-01-01

Results on parameter optimization for cybernetics model of RBMK reactor operator by power release control function are presented. Convolutions of various criteria applied previously in algorithms of the program 'Adviser to reactor operator' formed the basis of the model. 7 refs.; 4 figs

Economic incentives of short out-of-reactor time for fast breeder reactor fuel

International Nuclear Information System (INIS)

Bentley, B.W.; Haffner, D.R.

1975-01-01

Economic benefits (primarily reduced uranium ore and enrichment expenditures) can be realized by reducing the LMFBR out-of-reactor fuel cycle time only if process cost penalties and R and D costs can be minimized. The results of the evaluation presented show the potential gross benefits of reducing the out-of-reactor time and the effects of various associated cost penalties on these benefits. The gross benefit results estimate the potential savings in electrical power generation in the next 50 years using constant 1975 dollars and discounting the costs at 7 1 / 2 percent per year

Gas turbine modular helium reactor in cogeneration; Turbina de gas reactor modular con helio en cogeneracion

Energy Technology Data Exchange (ETDEWEB)

Leon de los Santos, G. [UNAM, Facultad de Ingenieria, Division de Ingenieria Electrica, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico, D. F. (Mexico)], e-mail: tesgleon@gmail.com

2009-10-15

This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

The safety basis of the integral fast reactor program

International Nuclear Information System (INIS)

Pedersen, D.R.; Seidel, B.R.

1990-01-01

The Integral Fast Reactor (IFR) and metallic fuel have emerged as the US Department of Energy reference reactor concept and fuel system for the development of an advanced liquid-metal reactor. This article addresses the basic elements of the IFR reactor concept and focuses on the safety advances achieved by the IFR Program in the areas of (1) fuel performance, (2) superior local faults tolerance, (3) transient fuel performance, (4) fuel-failure mechanisms, (5) performance in anticipated transients without scram, (6) core-melt mitigation, and (7) actinide recycle

RELAP-7 Software Verification and Validation Plan

Energy Technology Data Exchange (ETDEWEB)

Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk, Reliability, and Regulatory Support; Choi, Yong-Joon [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk, Reliability, and Regulatory Support; Zou, Ling [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk, Reliability, and Regulatory Support

2014-09-25

This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Reactor physics calculations in the Nordic countries

International Nuclear Information System (INIS)

Hoeglund, R.

1995-01-01

The seventh biennial meeting on reactor physics calculations in the Nordic countries was arranged by VTT Energy on May 8-9, 1995. 26 papers on different subjects in the field of reactor physics were presented by 45 participants representing research establishments, technical universities, utilities, consultants and suppliers. Resent development and verification of the program systems of ABB Atom, Risoe, Scandpower, Studsvik and VTT Energy were the main topic of the meeting. Benchmarking of the two assembly codes CASMO-4 and HELIOS is proceeding. Cross section data calculated with CASMO-HEX have been validated for the Loviisa reactors. On core analysis ABB atom gives a description on its latest core simulator version POLCA7 with the calculation Core Master 2 and the BWR core supervision system Core Watch. Transient calculations with HEXTRAN, HEXTRAN- PLIM, TRAB, RAMONA, SIMULATE-3K and a code based on PRESTO II/POLCA7 were also presented

RELAP-7 Closure Correlations

Energy Technology Data Exchange (ETDEWEB)

Zou, Ling [Idaho National Lab. (INL), Idaho Falls, ID (United States); Berry, R. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, R. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andrs, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hansel, J. E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sharpe, J. P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Johns, Russell C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-04-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework, MOOSE (Multi-Physics Object Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s and TRACE’s capabilities and extends their analysis capabilities for all reactor system simulation scenarios. The RELAP-7 code utilizes the well-posed 7-equation two-phase flow model for compressible two-phase flow. Closure models used in the TRACE code has been reviewed and selected to reflect the progress made during the past decades and provide a basis for the colure correlations implemented in the RELAP-7 code. This document provides a summary on the closure correlations that are currently implemented in the RELAP-7 code. The closure correlations include sub-grid models that describe interactions between the fluids and the flow channel, and interactions between the two phases.

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

Fusion reactor safety

International Nuclear Information System (INIS)

1987-12-01

Nuclear fusion could soon become a viable energy source. Work in plasma physics, fusion technology and fusion safety is progressing rapidly in a number of Member States and international collaboration continues on work aiming at the demonstration of fusion power generation. Safety of fusion reactors and technological and radiological aspects of waste management are important aspects in the development and design of fusion machines. In order to provide an international forum to review and discuss the status and the progress made since 1983 in programmes related to operational safety aspects of fusion reactors, their waste management and decommissioning concepts, the IAEA had organized the Technical Committee on ''Fusion Reactor Safety'' in Culham, 3-7 November 1986. All presentations of this meeting were divided into four sessions: 1. Statements on National-International Fusion Safety Programmes (5 papers); 2. Operation and System Safety (15 papers); 3. Waste Management and Decommissioning (5 papers); 4. Environmental Impacts (6 papers). A separate abstract was prepared for each of these 31 papers. Refs, figs, tabs

Inherent safety characteristics of innovative reactors

International Nuclear Information System (INIS)

Heil, J.A.

1995-11-01

of the considered innovative designs with passive safety features and increased simplicity, the modular gas-cooled reactors offer the best option for support. 5 figs., 4 tabs., 11 refs., 7 appendices

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

Innovative energy production in fusion reactors

International Nuclear Information System (INIS)

Iiyoshi, A.; Momota, H.; Motojima, O.

1994-01-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author)

Innovative energy production in fusion reactors

International Nuclear Information System (INIS)

Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

1993-10-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author)

Innovative energy production in fusion reactors

Science.gov (United States)

Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

1993-10-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are: (1) traveling wave direct energy conversion of 14.7 MeV protons; (2) cusp type direct energy conversion of charged particles; (3) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas; and (4) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising.

Innovative energy production in fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

1993-10-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author).

Effective utilization and management of research reactors

Energy Technology Data Exchange (ETDEWEB)

Muranaka, R [International Atomic Energy Agency, Vienna (Austria). Div. of Research and Isotopes

1984-06-01

The problem of utilizing a research reactor effectively is closely related to its management and therefore should not be considered separately. Too often, attention has been focused on specific techniques and methods rather than on the overall programme of utilization, with the result that skills and equipment have been acquired without any active continuing programme of applications and services. The seminar reported here provided a forum for reactor managers, users, and operators to discuss their experience. At the invitation of the Government of Malaysia, it was held at the Asia Pacific Development Centre, Kuala Lumpur, from 7 to 11 November 1983. It was attended by about 50 participants from 19 Member States; it is hoped that a report on the seminar, including papers presented, can be published and thus reach a wider audience. Thirty-one lectures and contributions were presented at a total of seven sessions: Research reactor management; Radiation exposure and safety; Research reactor utilization (two sessions); PUSPATI Research Reactor Project Development; Core conversion to low-enriched uranium, and safeguards; Research reactor technology. In addition, a panel discussed the causes and resolutions of the under-utilization of research reactors.

Effective utilization and management of research reactors

International Nuclear Information System (INIS)

Muranaka, R.

1984-01-01

The problem of utilizing a research reactor effectively is closely related to its management and therefore should not be considered separately. Too often, attention has been focused on specific techniques and methods rather than on the overall programme of utilization, with the result that skills and equipment have been acquired without any active continuing programme of applications and services. The seminar reported here provided a forum for reactor managers, users, and operators to discuss their experience. At the invitation of the Government of Malaysia, it was held at the Asia Pacific Development Centre, Kuala Lumpur, from 7 to 11 November 1983. It was attended by about 50 participants from 19 Member States; it is hoped that a report on the seminar, including papers presented, can be published and thus reach a wider audience. Thirty-one lectures and contributions were presented at a total of seven sessions: Research reactor management; Radiation exposure and safety; Research reactor utilization (two sessions); PUSPATI Research Reactor Project Development; Core conversion to low-enriched uranium, and safeguards; Research reactor technology. In addition, a panel discussed the causes and resolutions of the under-utilization of research reactors

Peer evaluation and some valuable lessons

Energy Technology Data Exchange (ETDEWEB)

Holt, A G [Ontario Hydro, Toronto, ON (Canada)

1991-04-01

In the mid 1980s there were some signs that Ontario Hydro's nuclear program performance was deteriorating. Such signs included increased maintenance backlog, increased number of jumpers, decreased capacity factors and increasing regulatory concerns. Factors influencing this deterioration were: (a) Pressure tube creep and hydriding rates were excessive leading to increased reactor maintenance and early pressure tube replacement in Pickering NGS-A and Bruce NGS-A. (b) Preventive maintenance was reduced to a minimum owing to manpower and budget restraints. This led to more forced outages, deratings and breakdown maintenance as the urgent was dealt with rather than the important. (c) New systems were installed in the older units, Pickering NGS-A and Bruce NGS-A, in order to backfit safety related system improvements principally to meet increased regulatory requirements. This put additional strain on tight resources to assist with the installation, commissioning, testing and maintenance of these systems that generally increased the complexity of units. Again this led to a reduction of preventive maintenance.

Peer evaluation and some valuable lessons

International Nuclear Information System (INIS)

Holt, A.G.

1991-01-01

In the mid 1980s there were some signs that Ontario Hydro's nuclear program performance was deteriorating. Such signs included increased maintenance backlog, increased number of jumpers, decreased capacity factors and increasing regulatory concerns. Factors influencing this deterioration were: (a) Pressure tube creep and hydriding rates were excessive leading to increased reactor maintenance and early pressure tube replacement in Pickering NGS-A and Bruce NGS-A. (b) Preventive maintenance was reduced to a minimum owing to manpower and budget restraints. This led to more forced outages, deratings and breakdown maintenance as the urgent was dealt with rather than the important. (c) New systems were installed in the older units, Pickering NGS-A and Bruce NGS-A, in order to backfit safety related system improvements principally to meet increased regulatory requirements. This put additional strain on tight resources to assist with the installation, commissioning, testing and maintenance of these systems that generally increased the complexity of units. Again this led to a reduction of preventive maintenance

Processes influencing cooling of reactor effluents

International Nuclear Information System (INIS)

Magoulas, V.E.; Murphy, C.E. Jr.

1982-01-01

Discharge of heated reactor cooling water from SRP reactors to the Savannah River is through sections of stream channels into the Savannah River Swamp and from the swamp into the river. Significant cooling of the reactor effluents takes place in both the streams and swamp. The majority of the cooling is through processes taking place at the surface of the water. The major means of heat dissipation are convective transfer of heat to the air, latent heat transfer through evaporation and radiative transfer of infrared radiation. A model was developed which incorporates the effects of these processes on stream and swamp cooling of reactor effluents. The model was used to simulate the effect of modifications in the stream environment on the temperature of water flowing into the river. Environmental effects simulated were the effect of changing radiant heat load, the effect of changes in tree canopy density in the swamp, the effect of total removal of trees from the swamp, and the effect of diverting the heated water from L reactor from Steel Creek to Pen Branch. 6 references, 7 figures

Os catadores de lixo e o processo de emancipação social Waste material pickers and emancipation process

Directory of Open Access Journals (Sweden)

Marta Pimenta Velloso

2005-12-01

Full Text Available O artigo tece comentários sobre o processo de organização de duas associações de catadores de materiais recicláveis. Traz à tona a problemática da exclusão, enfatizando a necessidade de inserção dos segmentos sociais marginalizados. Focaliza, como caminho de inserção, a recriação de espaços com ambiente apropriado ao desenvolvimento da criatividade e, conseqüentemente, ao processo de emancipação social. Apresenta o Pequeno Grupo como solo provisório para existência humana, onde o poder decisório se manifesta a partir da interação dinâmica entre o singular e o coletivo.The article describes the organizing process of two associations of waste and recyclings material pickers. The question of social exclusion emerges emphasizing therefore the urgent socially comeback of left out segments of people. It shows that the way to resocializing is the creation of environmentally approprieted spaces to develop creativity leading thus to self emancipation. The small group is seen as a transitory ground of human life, where the power to decide comes out as an interplay of individuals and collectives bodies.

Joint Effects of Granule Size and Degree of Substitution on Octenylsuccinated Sweet Potato Starch Granules As Pickering Emulsion Stabilizers.

Science.gov (United States)

Li, Jinfeng; Ye, Fayin; Lei, Lin; Zhou, Yun; Zhao, Guohua

2018-05-02

The granules of sweet potato starch were size fractionated into three portions with significantly different median diameters ( D 50 ) of 6.67 (small-sized), 11.54 (medium-sized), and 16.96 μm (large-sized), respectively. Each portion was hydrophobized at the mass-based degrees of substitution (DS m ) of approximately 0.0095 (low), 0.0160 (medium), and 0.0230 (high). The Pickering emulsion-stabilizing capacities of modified granules were tested, and the resultant emulsions were characterized. The joint effects of granule size and DS m on emulsifying capacity (EC) were investigated by response surface methodology. For small-, medium-, and large-sized fractions, their highest emulsifying capacities are comparable but, respectively, encountered at high (0.0225), medium (0.0158), and low (0.0095) DS m levels. The emulsion droplet size increased with granule size, and the number of freely scattered granules in emulsions decreased with DS m . In addition, the term of surface density of the octenyl succinic group (SD -OSG ) was first proposed for modified starch granules, and it was proved better than DS m in interpreting the emulsifying capacities of starch granules with varying sizes. The present results implied that, as the particulate stabilizers, the optimal DS m of modified starch granules is size specific.

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

Maintenance and material aspects of DREAM reactor

International Nuclear Information System (INIS)

Ueda, S.; Nishio, S.; Yamada, R.; Seki, Y.; Kurihara, R.; Adachi, J.; Yamazaki, S.

2000-01-01

A concept of a commercial fusion power reactors (Fusion Power: 5.5 GW, electric output: 2.7 GW) having high environmental safety, high thermal efficiency and high availability has been studied in JAERI. The gross reactor configuration was designed to achieve good maintainability, high performance breeding blanket, high efficient power generation system and little radwastes. Design was based on the use of low activation structural material (SiC/SiC composites) and helium as a coolant. In this paper, maintenance and material aspects of DREAM reactor design is discussed. The concluding remarks are as follows. (1) The difficulty of development of maintenance tool is alleviated by sector replacement and the radiation dose environment less than 10 Gy/h in a reactor chamber. (2) Design requirement and present status of SiC/SiC composites was investigated. (3) The SiC/SiC composite development program is planned to satisfy the requirements of DREAM reactor

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-03-01

A conceptual design study (option C) has been carried out for the fusion experimental reactor (FER). In addition to design of the tokamak reactor and associated systems based on the reference design specifications, feasibility of a water-shield reactor concept was examined as a topical study. The design study for the reference tokamak reactor has produced a reactor concept for the FER, along with major R D items for the concept, based on close examinations on thermal design, electromagnetics, neutronics and remote maintenance. Particular efforts have been directed to the area of electromagnetics. Detailed analyses with close simulation models have been performed on PF coil arrangements and configurations, shell effects of the blanket for plasma position unstability, feedback control, and eddy currents during disruptions. The major design specifications are as follows; Peak fusion power 437 MW Major radius 5.5 m Minor radius 1.1 m Plasma elongation 1.5 Plasma current 5.3 MA Toroidal beta 4 % Field on axis 5.7 T (author)

Effects of cryogenic reactor irradiation on organic insulators

International Nuclear Information System (INIS)

Kato, Teruo

1995-01-01

Insulators for the superconducting magnets of fusion reactor are classified as electrical and thermal insulators for which tough organic materials will be used. When the magnet is exposed by fast neutrons and gamma-rays from plasma in a fusion reactor, the fusion reactor systems will cause fatal damage by the degradation of insulators. Therefore, it is necessary to select materials resistant irradiation damage for use as insulators. Electrical and mechanical tests were carried out at 4.2 K without warmup after the reactor irradiation at 5 K. The effects of reactor irradiation at the dose of 10 7 Gy on epoxy resins (bisphenol-A), G-10 CR, VL-E 200 and G-11 CR caused large decreases in mechanical strength. Polyetheretherketone (PEEK), polyimide and phenol novolac resins, which were used to laminate reinforced plastics with glass-cloth against irradiation, showed good resistance. Effects of cryogenic reactor irradiation on several organic materials and epoxy laminate-reinforced plastics with glass-cloth and Kevlar-cloth were also discussed. (author)

FEM analysis of mechanical behaviour of coil support connections in Wendelstein 7-X fusion reactor

International Nuclear Information System (INIS)

Krzesinski, G.; Zagrajek, T.; Marek, P.; Dobosz, R.; Czarkowski, P.; Kurzydlowski, K.J.

2006-01-01

The objective of Wendelstein 7-X project is the stellarator-type fusion reactor. In this device plasma channel is under control of magnetic field coming from magnet system of very complicated shape, made of 70 superconducting coils symmetrically arranged in 5 identical sections. Every coil is connected to central ring with two extensions which transfer loads resulting from electromagnetic field and gravity. The aim of this work was to analyse mechanical behaviour of the bolted connections using detailed 3D finite element models. All simulations were performed assuming elasto-plastic behaviour of the materials, assembly stresses and friction contacts between different parts of the connections. Stress distributions, displacements, forces acting on the bolts and welds were studied using standard and submodeling routines. The results were subsequently used to optimize the design of critical central support elements. (author)

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

2004-01-01

Full text: The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

2004-01-01

The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance in the

ANAEROBIC DIGESTION AND THE DENITRIFICATION IN UASB REACTOR

Directory of Open Access Journals (Sweden)

José Tavares de Sousa

2008-01-01

Full Text Available The environmental conditions in Brazil have been contributing to the development of anaerobic systems in the treatment of wastewaters, especially UASB - Upflow Anaerobic Sludge Blanket reactors. The classic biological process for removal of nutrients uses three reactors - Bardenpho System, therefore, this work intends an alternative system, where the anaerobic digestion and the denitrification happen in the same reactor reducing the number of reactors for two. The experimental system was constituted by two units: first one was a nitrification reactor with 35 L volume and 15 d of sludge age. This system was fed with raw sanitary waste. Second unit was an UASB, with 7.8 L and 6 h of hydraulic detention time, fed with ¾ of effluent nitrification reactor and ¼ of raw sanitary waste. This work had as objective to evaluate the performance of the UASB reactor. In terms of removal efficiency, of bath COD and nitrogen, it was verified that the anaerobic digestion process was not affected. The removal efficiency of organic material expressed in COD was 71%, performance already expected for a reactor of this type. It was also observed that the denitrification process happened; the removal nitrate efficiency was 90%. Therefore, the denitrification process in reactor UASB is viable.

Experience in operation of heavy water reactors

International Nuclear Information System (INIS)

Rotaru, Ion; Bilegan, Iosif; Ghitescu, Petre

1999-01-01

The paper presents the main topics of the CANDU owners group (COG) meeting held in Mangalia, Romania on 7-10 September 1998. These meetings are part of the IAEA program for exchange of information related mainly to CANDU reactor operation safety. The first meeting for PHWR reactors took place in Vienna in 1989, followed by those in Argentina (1991), India (1994) and Korea (1996). The topics discussed at the meeting in Romania were: operation experience and recent major events, performances of CANDU reactors and safe operation, nuclear safety and operation procedures of PHWR, programs and strategies of lifetime management of installations and components of NPPs, developments and updates

Medium temperature carbon dioxide gas turbine reactor

International Nuclear Information System (INIS)

Kato, Yasuyoshi; Nitawaki, Takeshi; Muto, Yasushi

2004-01-01

A carbon dioxide (CO 2 ) gas turbine reactor with a partial pre-cooling cycle attains comparable cycle efficiencies of 45.8% at medium temperature of 650 deg. C and pressure of 7 MPa with a typical helium (He) gas turbine reactor of GT-MHR (47.7%) at high temperature of 850 deg. C. This higher efficiency is ascribed to: reduced compression work around the critical point of CO 2 ; and consideration of variation in CO 2 specific heat at constant pressure, C p , with pressure and temperature into cycle configuration. Lowering temperature to 650 deg. C provides flexibility in choosing materials and eases maintenance through the lower diffusion leak rate of fission products from coated particle fuel by about two orders of magnitude. At medium temperature of 650 deg. C, less expensive corrosion resistant materials such as type 316 stainless steel are applicable and their performance in CO 2 have been proven during extensive operation in AGRs. In the previous study, the CO 2 cycle gas turbomachinery weight was estimated to be about one-fifth compared with He cycles. The proposed medium temperature CO 2 gas turbine reactor is expected to be an alternative solution to current high-temperature He gas turbine reactors

Evaluation of the passage of Lactobacillus gasseri K7 and bifidobacteria from the stomach to intestines using a single reactor model

Directory of Open Access Journals (Sweden)

von Ah Ueli

2009-05-01

Full Text Available Abstract Background Probiotic bacteria are thought to play an important role in the digestive system and therefore have to survive the passage from stomach to intestines. Recently, a novel approach to simulate the passage from stomach to intestines in a single bioreactor was developed. The advantage of this automated one reactor system was the ability to test the influence of acid, bile salts and pancreatin. Lactobacillus gasseri K7 is a strain isolated from infant faeces with properties making the strain interesting for cheese production. In this study, a single reactor system was used to evaluate the survival of L. gasseri K7 and selected bifidobacteria from our collection through the stomach-intestine passage. Results Initial screening for acid resistance in acidified culture media showed a low tolerance of Bifidobacterium dentium for this condition indicating low survival in the passage. Similar results were achieved with B. longum subsp. infantis whereas B. animalis subsp. lactis had a high survival. These initial results were confirmed in the bioreactor model of the stomach-intestine passage. B. animalis subsp. lactis had the highest survival rate (10% attaining approximately 5 × 106 cfu ml-1 compared to the other tested bifidobacteria strains which were reduced by a factor of up to 106. Lactobacillus gasseri K7 was less resistant than B. animalis subsp. lactis but survived at cell concentrations approximately 1000 times higher than other bifidobacteria. Conclusion In this study, we were able to show that L. gasseri K7 had a high survival rate in the stomach-intestine passage. By comparing the results with a previous study in piglets we could confirm the reliability of our simulation. Of the tested bifidobacteria strains, only B. animalis subsp. lactis showed acceptable survival for a successful passage in the simulation system.

Proceedings of the sixth Asian symposium on research reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-08-01

The symposium consisted of 16 sessions with 58 submitted papers. Major fields were: (1) status and future plan of research and testing reactors, (2) operating experiences, (3) design and modification of the facility, and reactor fuels, (4) irradiation studies, (5) irradiation facilities, (6) reactor characteristics and instrumentation, and (7) neutron beam utilization. Panel discussion on the 'New Trends on Application of Research and Test Reactors' was also held at the last of the symposium. About 180 people participated from China, Korea, Indonesia, Thailand, Bangladesh, Vietnam, Chinese Taipei, Belgium, France, USA, Japan and IAEA. The 58 of the presented papers are indexed individually. (J.P.N.)

Proceedings of the sixth Asian symposium on research reactors

International Nuclear Information System (INIS)

1999-08-01

The symposium consisted of 16 sessions with 58 submitted papers. Major fields were: 1) status and future plan of research and testing reactors, 2) operating experiences, 3) design and modification of the facility, and reactor fuels, 4) irradiation studies, 5) irradiation facilities, 6) reactor characteristics and instrumentation, and 7) neutron beam utilization. Panel discussion on the 'New Trends on Application of Research and Test Reactors' was also held at the last of the symposium. About 180 people participated from China, Korea, Indonesia, Thailand, Bangladesh, Vietnam, Chinese Taipei, Belgium, France, USA, Japan and IAEA. The 58 of the presented papers are indexed individually. (J.P.N.)

MIT research reactor. Power uprate and utilization

Energy Technology Data Exchange (ETDEWEB)

Hu, Lin-Wen [Nuclear Reactor Laboratory, Massachusetts Inst. of Technology, Cambridge, MA (United States)

2012-03-15

The MIT Research Reactor (MITR) is a university research reactor located on MIT campus. and has a long history in supporting research and education. Recent accomplishments include a 20% power rate to 6 MW and expanding advanced materials fuel testing program. Another important ongoing initiative is the conversion to high density low enrichment uranium (LEU) monolithic U-Mo fuel, which will consist of a new fuel element design and power increase to 7 MW. (author)

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

2003-01-01

Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

Muon trackers for imaging a nuclear reactor

Science.gov (United States)

Kume, N.; Miyadera, H.; Morris, C. L.; Bacon, J.; Borozdin, K. N.; Durham, J. M.; Fuzita, K.; Guardincerri, E.; Izumi, M.; Nakayama, K.; Saltus, M.; Sugita, T.; Takakura, K.; Yoshioka, K.

2016-09-01

A detector system for assessing damage to the cores of the Fukushima Daiichi nuclear reactors by using cosmic-ray muon tomography was developed. The system consists of a pair of drift-tube tracking detectors of 7.2× 7.2-m2 area. Each muon tracker consists of 6 x-layer and 6 y-layer drift-tube detectors. Each tracker is capable of measuring muon tracks with 12 mrad angular resolutions, and is capable of operating under 50-μ Sv/h radiation environment by removing gamma induced background with a novel time-coincidence logic. An estimated resolution to observe nuclear fuel debris at Fukushima Daiichi is 0.3 m when the core is imaged from outside the reactor building.

Managing severe reactor accidents. A review and evaluation of our knowledge on reactor accidents and accident management

International Nuclear Information System (INIS)

Gustavsson, Veine

2002-11-01

The report gives a review of the results from the last years research on severe reactor accidents, and an opinion on the possibilities to refine the present strategies for accident management in Swedish and Finnish BWRs. The following aspect of reactor accidents are the major themes of the study: 1. Early pressure relief from hydrogen production; 2. Recriticality in re-flooded, degraded core; 3. Melt-through; 4. Steam explosion after melt-through; 5. Coolability of the melt after after melt-through; 6. Hydrogen fire in the reactor containment; 7. Leaking containment; 8. Hydrogen fire in the reactor building; 9. Long-time developments after a severe accident; 10. Accidents during shutdown for overhaul; 11. Information need for remedial actions. Possibilities for improving the strategies in each of these areas are discussed. The review shows that our knowledge is sufficient in the areas 1, 2, 4, 6, 8. For the other areas, more research is needed

Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor; Ispitivanje reaktorskih parametara na kriticnim sistemima, I faza: Izvestaj o sigurnosti reaktora nulte snage RB

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

1962-09-15

In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined.

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

Science.gov (United States)

Clement, J. D.; Rust, J. H.

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

A remotely controlled CCTV system for nuclear reactor retube operations

International Nuclear Information System (INIS)

Stovman, J.A.

1984-01-01

This paper describes the CCTV Vault Observation Subsystem (VOS) under development for Ontario Hydro for the Pickering 'A' Nuclear Power Plant Large Scale Retubing program. This subsystem will be used by a supervisor and several operators to observe fuel channel replacement operations following plant shutdown and removal of the fuel bundles. VOS basically comprises 23 monochrome television camera driven circuits, a matrix switcher, 15 monitors, 9 tape recorders and 4 microphone driven sound circuits. Remote control of the camera's zoom lenses and mounts is via a digitally multiplexed control system. Design considerations include viewing requirements, reliability, radiation, redundance, and economic factors

Licensed operating reactors status summary report data as of April 30, 1983. Vol. 7, No. 5

International Nuclear Information System (INIS)

1983-05-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors: status summary report, data as of May 31, 1983. Volume 7, No. 6

International Nuclear Information System (INIS)

1983-06-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

International Nuclear Information System (INIS)

Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

2009-01-01

A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m 3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m 3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m 3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m 3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m 3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m 3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

Luncheon address: Development of the CANDU reactor

International Nuclear Information System (INIS)

Bain, A.S.

1997-01-01

The paper is a highlight of the some of the achievements in the development of the CANDU Reactor, taken from the book C anada Enters the Nuclear Age . The CANDU reactor is one of Canada's greatest scientific/engineering achievements, that started in the 1940's and bore fruit with the reactors of the 60's, 70's, and 80's. The Government decided in the 1950's to proceed with a demonstration nuclear power reactor (NPD), AECL invited 7 Canadian corporations to bid on a contract to design and construct the NPD plant. General Electric was selected. A utility was also essential for participation and Ontario Hydro was chosen. In May 1957 it was concluded that the minimum commercial size would be about 200MWe and it should use horizontal pressure tubes to contain the fuel and pressurized heavy water coolant. The book also talks of standard out-reactor components such as pumps, valves, steam generators and piping. A major in-reactor component of interest was the fuel, fuel channels and pressure tubes. A very high level of cooperation was required for the success of the CANDU program

The influence of Triga 2000 reactor operation on the surface contamination at reactor room using smear test method

International Nuclear Information System (INIS)

Bintu Khoiriyyah; Budi Purnama; Tri Cahyo Laksono

2016-01-01

The monitoring of surface contamination should be conducted to determine the safety of work areas. Surface contamination at the TRIGA 2000 reactor room which is on PSTNT-BATAN Bandung remain to be implemented although reactor not operating. In this research monitoring of surface contamination when TRIGA 2000 in operation of the first time after several years not operating aims to determine the influence on the results of monitoring. The monitoring of surface contamination has been done using smear test method at some predetermined in TRIGA 2000 reactor room. The highest surface contamination activities is obtained 0.32 Bq/cm 2 and there are some points that are not detected. Based on keputusan kepala BAPETEN No.1/Ka BAPETEN/ V/99 the work showed that the TRIGA 2000 reactor in the category of low area contamination, that is <3.7 Bq/cm 2 to gross beta. (author)

Quality assurance in the manufacture of metallic uranium fuel for research reactors

International Nuclear Information System (INIS)

Shah, B.K.; Kumar, Arbind; Nanekar, P.P.; Vaidya, P.R.

2009-01-01

Two Research Reactors viz. CIRUS and DHRUVA are operating at Trombay since 1960 and 1985 respectively. Cirus is a 40 MWth reactor using heavy water as moderator and light water as coolant. Dhruva is a 100 MWth reactor using heavy water as moderator and coolant. The maximum neutron flux of these reactors are 6.7 x 10 13 n/cm 2 /s (Cirus) and 1.8 x 10 14 n/cm 2 /s (Dhruva). Both these reactors are used for basic research, R and D in reactor technology, isotope production and operator training. Fuel material for these reactors is natural uranium metallic rods claded in finned aluminium (99.5%) tubes. This presentation will discuss various issues related to fabrication quality assurance and reactor behavior of metallic uranium fuel used in research reactors

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

Fabrication of ORNL Fuel Irradiated in the Peach Bottom Reactor and Postirradiation Examination of Recycle Test Elements 7 and 4

International Nuclear Information System (INIS)

Long, Jr. E.L.

2001-01-01

Seven full-sized Peach Bottom Reactor fuel elements were fabricated in a cooperative effort by Oak Ridge National Laboratory (ORNL) and Gulf General Atomic (GGA) as part of the National HTGR Fuel Recycle Development Program. These elements contain bonded fuel rods and loose beds of particles made from several combinations of fertile and fissile particles of interest for present and future use in the High-Temperature Gas-Cooled Reactor (HTGR). The portion of the fuel prepared for these elements by ORNL is described in detail in this report, and it is in conjunction with the GGA report (GA-10109) a complete fabrication description of the test. In addition, this report describes the results obtained to date from postirradiation examination of the first two elements removed from the Peach Bottom Reactor, RTE-7 and -4. The fuel examined had relatively low exposure, up to about 1.5 x 10 21 neutrons/cm* fast (>0.18 MeV) fluence, compared with the peak anticipated HTGR fluence of 8.0 x 10 21 , but it has performed well at this exposure. Dimensional data indicate greater irradiation shrinkage than expected from accelerated test data to higher exposures. This suggests that either the method of extrapolation of the higher exposure data back to low exposure is faulty, or the behavior of the coated particles in the neutron spectrum characteristic of the accelerated tests does not adequately represent the behavior in an HTGR spectrum

Phosphorus removal from UASB reactor effluent by reactive media filtration.

Science.gov (United States)

Rodríguez-Gómez, Raúl; Renman, Gunno

2017-08-01

The phosphorus (P) and BOD7 removal performance of an upflow packed bed reactor (PBR) filled with two reactive filter media was studied over 50 weeks. The lower one-fifth of the reactor was filled with calcium-silicate-hydrate (Sorbulite®) and the upper four-fifth with calcium-silicate (Polonite®). A laboratory-scale upflow anaerobic sludge bed reactor (UASB) delivered wastewater to the PBR. A model was developed to describe the gradient in P concentration change in the reactor, based on reaction kinetics. The reaction terms were assumed to follow the Langmuir isotherm, based on the results obtained in a batch test. First, a comparison was made between experimental and simulated results. The capability of the model to forecast P removal capacity was then tested for three hypothetical cases: (i) reactor filled with Sorbulite and Polonite, (ii) reactor filled with only Sorbulite, and (iii) reactor filled with only Polonite. Finally, a sensitivity analysis was performed for the main parameters in the model. The average removal of P and BOD7 from the UASB effluent was 98% and 90%, respectively. The starting pH of the dual-medium effluent was 12.2 and decreased gradually over time to 11.1. The simulation both overestimated and underestimated mean measured P removal but was within the range of maximum and minimum measured values. The hypothetical cases revealed that most P was removed by Polonite due to calcium phosphate precipitation. The removal capacity of the two filter materials and their layer height in the reactor were the most sensitive parameters in the simulation.

U.S. Status of Fast Reactor Research and Technology

International Nuclear Information System (INIS)

Hill, Robert

2012-01-01

Summary: • Fast reactor R&D is focused on key technologies innovations for performance improvement (cost reduction) and safety: 1. System Integration and Concept Development; 2. Safety Technology; 3. Advanced Materials; 4. Ultrasonic Viewing; 5. Advanced Energy Conversion (Supercritical CO 2 Brayton cycle); 6. Reactor Simulation; 7. Nuclear Data; 8. Advanced Fuels. • Fast reactors have flexible capability for actinide management: – A wide variety of fuel cycle options are being considered; • International R&D collaboration pursued in Generation-IV and multilateral arrangements

Analysis of an accelerator-driven subcritical light water reactor

International Nuclear Information System (INIS)

Kruijf, W.J.M. de; Wakker, P.H.; Wetering, T.F.H. van de; Verkooijen, A.H.M.

1997-01-01

An analysis of the basic characteristics of an accelerator-driven light water reactor has been made. The waste in the nuclear fuel cycle is considerably less than in the light water reactor open fuel cycle. This is mainly caused by the use of equilibrium nuclear fuel in the reactor. The accelerator enables the use of a fuel composition with infinite multiplication factor k ∞ < 1. The main problem of the use of this type of fuel is the strongly peaked flux distribution in the reactor core. A simple analytical model shows that a large core is needed with a high peak power factor in order to generate net electric energy. The fuel in the outer regions of the reactor core is used very poorly. 7 refs., 4 figs., 1 tab

Plugging inaccessible leaks in cooling water pipework in nuclear power plants

International Nuclear Information System (INIS)

Powell, A.B.; May, R.; Down, M.G.

1988-01-01

The manifestation of initially small leaks in ancilliary reactor cooling water systems is not an unusual event. Often these leaks are in virtually inaccessible locations - for example, buried in thick concrete shielding or situated in cramped and highly radioactive vaults. Such leaks may ultimately prejudice the availability of the entire nuclear system. Continued operation without repair can result in the leak becoming larger, and the leaking water can cause further corrosion problems and interfere with instrumentation. In addition, the water may increase the volume of radwaste. In short, initially trivial leaks may cause significant operating problems. This paper describes the sealing of such leaks in the biological shield cooling system of Ontario Hydro's Pickering nuclear generating station CANDU reactors

Neutron behavior, reactor control, and reactor heat transfer. Volume four

International Nuclear Information System (INIS)

Anon.

1986-01-01

Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

Gas turbine modular helium reactor in cogeneration

International Nuclear Information System (INIS)

Leon de los Santos, G.

2009-10-01

This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

Fast reactors - Dounreay and the future

International Nuclear Information System (INIS)

Jordan, G.

1988-01-01

In 1960 at Dounreay, the Dounreay Fast Reactor (DFR) supplied the world's first fast reactor grid electricity, and went on to a highly successful career as a test facility, as fuel designs advanced. In the 1960s, the Prototype Fast Reactor (PFR) was designed and built, beginning operation in 1974. The PFR was built to provide a sound technical and experienced base to support the UK's future Fast Reactor development and design. The in-vessel fuel handling facilities have demonstrated the flexibility of the pool design and a considerable body of in-core fuel handling experience is available. A key issue for further Fast Reactor application is the performance of fuel and, because PFR was designed to take full-scale fuel assemblies, the fuel performance experience is directly relevant to commercial designs. The original PFR design irradiation target of 60000 MWd/t U (equivalent to 7.5 % burn-up) has already been exceeded by a factor of more than two and a 15.9 % burn-up sub-assembly has been discharged and reprocessed without difficulty. Soon a 20 % sub-assembly will follow. Also the PFR reprocessing plant has demonstrated the safety and efficiency of this essential adjunct to Fast Reactor operation. The safety and the environmental protection features of both the PFR and its fuel reprocessing plant have been demonstrated over the last 14 years. 2 refs., 3 figs

New training reactor at Dresden Technical University

International Nuclear Information System (INIS)

Hansen, W.; Knorr, J.; Wolf, T.

2006-01-01

A total of 14 low-power (up to 10 W) training reactors have been operated at German universities, 9 of them officially classified as being operational in 2004, though for very different uses. This number is expected to drop sharply. The only comprehensive upgrading of a training reactor took place at Dresden Technical University: AKR-2, the most modern facility in Germany, started routine operation in April 2005, under a newly granted license pursuant to Sec. 7, Subsec. 1 of the German Atomic Energy Act, for training students in nuclear technology, for suitable research projects, and a a center of information about reactor technology and nuclear technology for the interested public. One special aspect of this refurbishment was the installation of digital safety I and C systems of the TELEPERM XS line, which are used also in other modern plants. This fact, plus the easy possibility to use the plant for many basic experiments in reactor physics and radiation protection, make the AKR-2 attractive also to other users (e.g. for training reactor personnel or other persons working in nuclear technology). (orig.)

ANALISIS PERMUKIMAN PEMULUNG SEBAGAI SEBUAH ASSEMBLAGE STUDI KASUS: PERMUKIMAN PEMULUNG DI WILAYAH JURANGMANGU, TANGERANG SELATAN

Directory of Open Access Journals (Sweden)

Eka Permanasari

2017-01-01

There are some research on social and economical condition of informal waste-picker settlements. However, the research on the urban assemblage of informal waste-picker settlements are not very common. Therefore, this research focus on how the social and economical condition shape the informal waste-picker settlements in terms of urban assemblage.   This research investigates the development and transformation of the informal waste-picker settlements in two prime locations in Jurangmangu, South Tangerang. The methods are through observation and interview to key persons (leaders on each settlement. The urban assemblage on these waste-picker settlements is heavily influenced by social and economical condition and activity of their users.   Key words: waste picker settlements, urban assemblage, informality

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

Energy Technology Data Exchange (ETDEWEB)

Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mueller, Don [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-12-01

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 ⁷LiF + BeF₂) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in ⁷Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.

Thermionic nuclear reactor systems

International Nuclear Information System (INIS)

Kennel, E.B.

1986-01-01

Thermionic nuclear reactors can be expected to be candidate space power supplies for power demands ranging from about ten kilowatts to several megawatts. The conventional ''ignited mode'' thermionic fuel element (TFE) is the basis for most reactor designs to date. Laboratory converters have been built and tested with efficiencies in the range of 7-12% for over 10,000 hours. Even longer lifetimes are projected. More advanced capabilities are potentially achievable in other modes of operation, such as the self-pulsed or unignited diode. Coupled with modest improvements in fuel and emitter material performance, the efficiency of an advanced thermionic conversion system can be extended to the 15-20% range. Advanced thermionic power systems are expected to be compatible with other advanced features such as: (1) Intrinsic subcritically under accident conditions, ensuring 100% safety upon launch abort; (2) Intrinsic low radiation levels during reactor shutdown, allowing manned servicing and/or rendezvous; (3) DC to DC power conditioning using lightweight power MOSFETS; and (4) AC output using pulsed converters

Reactor physics aspects of CANDU reactors

International Nuclear Information System (INIS)

Critoph, E.

1980-01-01

These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

Proceedings of the symposium on the physics and technology of reactors

International Nuclear Information System (INIS)

1993-01-01

The symposium aimed at providing the opportunity for promoting the subject and for developing the human resources in this important field in the Arab States. The symposium included 32 lectures on the following topics related to research reactors: design and development, training and operation, calculations of reactor parameters, nuclear reactions dynamics and control, reactor physics, neutron pyhsics, neutron activation analysis, in-core reactor radiation protection and shielding calculations. The lectures of the symposium were distributed over 7 sessions. An additional session was held by all participants for open discussion and recommendations

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

Peptidolytic microbial community of methanogenic reactors from two modified UASBs of brewery industries

Directory of Open Access Journals (Sweden)

C. Díaz

2010-10-01

Full Text Available We studied the peptide-degrading anaerobic communities of methanogenic reactors from two mesophilic full-scale modified upflow anaerobic sludge blanket (UASB reactors treating brewery wastewater in Colombia. Most probable number (MPN counts varied between 7.1 x 10(8 and 6.6 x 10(9 bacteria/g volatile suspended solids VSS (Methanogenic Reactor 1 and 7.2 x 10(6 and 6.4 x 10(7 bacteria/g (VSS (Methanogenic Reactor 2. Metabolites detected in the highest positive MPN dilutions in both reactors were mostly acetate, propionate, isovalerate and, in some cases, negligible concentrations of butyrate. Using the highest positive dilutions of MPN counts, 50 dominant strains were isolated from both reactors, and 12 strains were selected for sequencing their 16S rRNA gene based on their phenotypic characteristics. The small-subunit rRNA gene sequences indicated that these strains were affiliated to the families Propionibacteriaceae, Clostridiaceae and Syntrophomonadaceae in the low G + C gram-positive group and Desulfovibrio spp. in the class d-Proteobacteria. The main metabolites detected in the highest positive dilutions of MPN and the presence of Syntrophomonadaceae indicate the effect of the syntrophic associations on the bioconversion of these substrates in methanogenic reactors. Additionally, the potential utilization of external electron acceptors for the complete degradation of amino acids by Clostridium strains confirms the relevance of these acceptors in the transformation of peptides and amino acids in these systems.

Fast reactor development programme in France during 1995

International Nuclear Information System (INIS)

Le Rigoleur, C.

1996-01-01

In 1995, the total amount of electricity produced in France was 471 TWh, out of which 358.2 TWh (76 %) were produced by nuclear power plants, 36.9 TWh (7.8 %) by conventional thermal plants, and 75.5 TWh (16 %) by hydraulic plants. The net electrical power consumption was 368.7 TWh. At the end of 1995, 'Electricite de France' had 54 PWR units in operation. The availability factor for these units was maintained at 81%. 1995 was marked by a decrease of unexpected shutdowns (1.8% in 1995 instead of 2.2% in 1994), a new reduction in programmed shutdown periods, and a good safety level was maintained. In the field of Fast Reactors, the main events of 1995 were the following. At the end of December 1994, the PHENIX reactor was authorized to perform its 49th cycle at 350 MW th (143 MWe). This 49th cycle was completed without any significant problems on April 7, 1995. During the remainder of the year, the reactor had been shut down in order to carry out several tasks within the scope of the ten-year extension of the PHENIX reactor's lifetime. Concerning the CREYS-MALVILLE plant (SUPER-PHENIX) the first part of the year was devoted to repairing argon leak of one of the IHX. Authorization to restart the reactor was given on August 22. The end of the year was beset by a number of minor incidents. The reactor was restarted at the end of 1995 and reactor power was increased by successive steps (30% Pn (Nominal Power) up to February 6 1996; followed by 50 %...). The 'Decret d'Autorisation de Creation' stipulates that because of its prototype character, SUPER PHENIX will have to be operated under conditions explicitly giving priority to safety and knowledge acquisition, with an objective of research and demonstration. In this context, the so-called 'knowledge acquisition' programme designed to prove the capacity of a large FBR to produce electricity on an industrial scale, to test the consumption of plutonium and minor actinides in a large fast reactor, as well as to provide

Permeated defect detecting test method and device in reactor

International Nuclear Information System (INIS)

Sakurai, Yoshishige.

1996-01-01

The present invention provides a method of and a device capable of performing a test for entire inner surfaces of the reactor upon periodical inspection of a BWR type reactor while sufficiently taking countermeasures for radiation rays into consideration. Namely, the present invention comprises following steps. (1) A provisional step for taking a shroud head of a reactor core shroud and incore structural components above and below the shroud out of the reactor, discharging reactor water and water tightly closing openings such as reactor wall perforation holes, (2) a pretreatment step for washing exposed inner surfaces of the reactor and peeling deteriorated materials, (3) a first drying step for drying portions washed and peeled in the step (2), (4) a permeation step for applying a permeation liquid of a defect detecting medium on the exposed inner surfaces of the reactor, (5) a permeation liquid removing step for removing the an excess permeation liquid in the step (4), (6) a second drying step for drying corresponding portions after performing the step (5), and (7) a flaw detecting step for optically observing the corresponding portions after performing the step (6) and detecting flaws. (I.S.)

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2002-01-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2001-01-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

Reactor

International Nuclear Information System (INIS)

Toyama, Masahiro; Kasai, Shigeo.

1978-01-01

Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

A stability identification system for boiling water nuclear reactors

International Nuclear Information System (INIS)

Belblidia, L.A.; Chevrier, A.

1994-01-01

Boiling water reactors are subject to instabilities under low-flow, high-power operating conditions. These instabilities are a safety concern and it is therefore important to determine stability margins. This paper describes a method to estimate a measure of stability margin, called the decay ratio, from autoregressive modelling of time series data. A phenomenological model of a boiling water reactor with known stability characteristics is used to generate time series to validate the program. The program is then applied to signals from local power range monitors from the cycle 7 stability tests at the Leibstadt plant. (author) 7 figs., 2 tabs., 12 refs

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

7Q10 flows for SRS streams

International Nuclear Information System (INIS)

Chen, K.F.

1996-01-01

The Environmental Transport Group of the Environmental Technology Section was requested to predict the seven-day ten-year low flow (7Q10 flow) for the SRS streams based on historical stream flow records. Most of the historical flow records for the SRS streams include reactor coolant water discharged from the reactors and process water released from the process facilities. The most straight forward way to estimate the stream daily natural flow is to subtract the measured upstream reactor and/or facility daily effluents from the measured downstream daily flow. Unfortunately, this method does not always work, as indicated by the fact that sometimes the measured downstream volumetric flow rates are lower than the reactor effluent volumetric flow rates. For those cases that cannot be analyzed with the simple subtracting method, an alternative method was used to estimate the stream natural flows by statistically separating reactor coolant and process water flow data. The correlation between the calculated 7Q10 flows and the watershed areas for Four Mile Branch and Pen Branch agrees with that calculated by the USGS for Upper Three Runs and Lower Three Runs Creeks. The agreement between these two independent calculations lends confidence to the 7Q10 flow calculations presented in this report

Caramel fuel for research reactors

International Nuclear Information System (INIS)

Bussy, P.

1979-11-01

This fuel for research reactors is made of UO 2 pellets in a zircaloy cladding to replace 93% enriched uranium. It is a cold fuel, non contaminating and non proliferating, enrichment is only 7 to 8%. Irradiation tests were performed until burn-up of 50000 MWD/t [fr

Coil winding pack FE-analysis for a HELIAS reactor

International Nuclear Information System (INIS)

Schauer, F.; Egorov, K.; Bykov, V.

2011-01-01

At the Max-Planck-Institut fuer Plasmaphysik (IPP) a reference design is being created of an upgraded five-periodic HELIAS type stellarator reactor which evolves from Wendelstein 7-X (W7-X) by scaling of the coil centre line geometries by a factor of four. This reactor type was extensively investigated at IPP with regard to physical characteristics and to some extent also to engineering issues. The upgrade concerns an increase of the induction at the plasma axis and correspondingly at the superconductor. The aim is to develop the reactor concept to a stage and such detail that major engineering problems are unveiled, and relevant comparisons with other concepts, including tokamaks, can be drawn in view of upcoming decisions concerning a DEMO reactor. Even though progress in plasma physics, and in particular future results of W7-X and other machines - particularly of ITER - will probably lead to somewhat different coil shapes, no principal changes of the reference design are expected. In this paper the option of a roll-formed square coil cable jacket is investigated. Detailed structural FE analysis of the coil winding pack demonstrates the feasibility of such a conductor which appears to be the most economical option. It also allows sufficient space for a cable current density very similar to that of the ITER TF coil with a similar overall winding pack cross section of ∼0.5 m 2 . Already existing Nb 3 Sn conductors could thus be safely applied in such a HELIAS reactor. Obvious progress of superconductor technology, particularly concerning Nb 3 Al, will be beneficial concerning savings of conductor material, ease of manufacture, higher operation temperature, etc.

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

Results of the Nucifer reactor neutrino experiment

Energy Technology Data Exchange (ETDEWEB)

Buck, Christian; Lindner, Manfred [MPIK Heidelberg (Germany)

2016-07-01

Nuclear reactors are a strong and pure source of electron antineutrinos. With neutrino experiments close to compact reactor cores new insights into neutrino properties and reactor physics can be obtained. The Nucifer experiment is one of the pioneers in this class of very short baseline projects. Its detector to reactor distance is only about 7 m. The data obtained in the last years allowed to estimate the plutonium concentration in the reactor core by the neutrino flux measurement. This is of interest for safeguard applications and non proliferation efforts. The antineutrinos in Nucifer are detected via the inverse beta decay on free protons. Those Hydrogen nuclei are provided by 850 liters of organic liquid scintillator. For higher detection efficiency and background reduction the liquid is loaded with Gadolinium. Despite all shielding efforts and veto systems the background induced by the reactor activity and cosmogenic particles is still the main challenge in the experiment. The principle of the Nucifer detector is similar to the needs of upcoming experiments searching for sterile neutrinos. Therefore, the Nucifer results are also valuable input for the understanding and optimization of those next generation projects. The observation of sterile neutrinos would imply new physics beyond the standard model.

Statement to International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, 7 December 2009, Kyoto, Japan

International Nuclear Information System (INIS)

Amano, Yukiya

2009-01-01

Full text: Distinguished Guests, Ladies and Gentlemen, It is my honour to address participants at this opening session of the International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities, organized by the IAEA and hosted by the Japan Atomic Energy Agency. Fast reactor technology has the potential to ensure that energy resources which would last hundreds of years with the technology we are using today will actually last several thousand years. In other words it can withstand enormous increases in demand. This innovative technology also reduces the risk to the environment and helps to limit the burden that will be placed on future generations in the form of waste products. The coming year will be an exciting one for the development of fast-spectrum nuclear reactors. We expect to reach many important milestones: - the first criticality of the China Experimental Fast Reactor; - the restart of the Monju prototype fast reactor in Japan; and - the new insights we will gain through the end-of-life studies at the Phenix reactor in France. In the near future, new fast reactors will be commissioned: the 500MW(e) Prototype Fast Breeder Reactor in India, the first in a series of five of the same type, and the BN-800 reactor in the Russian Federation. Moreover, France, Japan, India, China and the Republic of Korea are preparing advanced prototypes, demonstration or commercial reactors for the 2020-2030 period. Nuclear power is set to be an increasingly important part of the global energy mix in the coming decades as demand for energy grows. Scores of countries in both the developed and developing world have told the IAEA that they are interested in introducing nuclear power. The 30 countries which already have nuclear power reactors are set to build more. This trend is likely to be accompanied by accelerated deployment of fast reactors. Continued advances in research and technology development are necessary to ensure improved economics and

Reactor operation

CERN Document Server

Shaw, J

2013-01-01

Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

Near-surface traveltime tomographic inversion using multiple first break picks

KAUST Repository

Saragiotis, Christos; Choi, Yun Seok; Keho, T.; Alkhalifah, Tariq Ali

2013-01-01

The input data for refraction traveltime tomography are the traveltimes of the first breaks, which are picked using automatic pickers. Although automatic pickers perform satisfactorily overall, no one automatic picker can be characterized as the best one; one picker might fail for traces for which other pickers are accurate and vice versa for other traces. We introduce an iterative method for traveltime tomography, which takes as input traveltimes from a number of pickers. During the inversion scheme inconsistent traveltimes are replaced with more meaningful ones to obtain a smooth near-surface velocity model. The scheme is easily parallelizable and a byproduct of the inversion scheme is a set of consistent traveltimes which is close to the actual traveltimes of the first breaks.

Satisfaction of patients hospitalised in psychiatric hospitals: a randomised comparison of two psychiatric-specific and one generic satisfaction questionnaires

Directory of Open Access Journals (Sweden)

Cléopas Agatta

2006-08-01

Full Text Available Abstract Background While there is interest in measuring the satisfaction of patients discharged from psychiatric hospitals, it might be important to determine whether surveys of psychiatric patients should employ generic or psychiatry-specific instruments. The aim of this study was to compare two psychiatric-specific and one generic questionnaires assessing patients' satisfaction after a hospitalisation in a psychiatric hospital. Methods We randomised adult patients discharged from two Swiss psychiatric university hospitals between April and September 2004, to receive one of three instruments: the Saphora-Psy questionnaire, the Perceptions of Care survey questionnaire or the Picker Institute questionnaire for acute care hospitals. In addition to the comparison of response rates, completion time, mean number of missing items and mean ceiling effect, we targeted our comparison on patients and asked them to answer ten evaluation questions about the questionnaire they had just completed. Results 728 out of 1550 eligible patients (47% participated in the study. Across questionnaires, response rates were similar (Saphora-Psy: 48.5%, Perceptions of Care: 49.9%, Picker: 43.4%; P = 0.08, average completion time was lowest for the Perceptions of Care questionnaire (minutes: Saphora-Psy: 17.7, Perceptions of Care: 13.7, Picker: 17.5; P = 0.005, the Saphora-Psy questionnaire had the largest mean proportion of missing responses (Saphora-Psy: 7.1%, Perceptions of Care: 2.8%, Picker: 4.0%; P P Conclusion Despite differences in the intended target population, content, lay-out and length of questionnaires, none appeared to be obviously better based on our comparison. All three presented advantages and drawbacks and could be used for the satisfaction evaluation of psychiatric inpatients. However, if comparison across medical services or hospitals is desired, using a generic questionnaire might be advantageous.

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

Reactor as furnace and reactor as lamp

International Nuclear Information System (INIS)

Goldanskii, V.I.

1992-01-01

There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Nufarhana Ayuni Joha; Mohd Sabri Minhat

2012-01-01

The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

Tritium release during inspection of reactor 'RA' at 'Vinca' institute

International Nuclear Information System (INIS)

Sipka, V.; Miljevic, N.; Grsic, Z.; Todorovic, D.; Radenkovic, M.

1997-01-01

Tritium content in daily precipitation, atmospheric water vapor inside of the reactor hall and around 'Vinca' Institute as well as in soil up to 800 m distance was monitoring during the regular inspection of the fuel channels. Tritium activity in the reactor hall air moisture was in the range from 0.022 to 6.7 MBq/m 3 . Tritium content in soil moisture between 12.7 and 530.9 Bq/l indicate a certain contamination due to tritium release in the environment, depending on the depth and distance from the place of release (author) [sr

Reactor safeguards

CERN Document Server

Russell, Charles R

1962-01-01

Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

1998-01-01

To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

Nuclear reactors for space electric power

International Nuclear Information System (INIS)

Buden, D.

1978-06-01

The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Lifetimes of 7 to 10 yr at full power, at converter operating temperatures of 1275 to 1675 0 K, are being studied. The systems are being designed such that no single-failure modes exist that will cause a complete loss of power. In fact, to meet the long lifetimes, highly redundant design features are being emphasized. Questions have been raised about safety since the COSMOS 954 incident. ''Fail-safe'' means to prevent exposure of the population to radioactive material, meeting the environmental guidelines established by the U.S. Government have been and continue to be a necessary requirement for any space reactor program. The major safety feature to prevent prelaunch and launch radioactive material hazards is not operating the reactor before achieving the prescribed orbit. Design features in the reactor ensure that accidental criticality cannot occur. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit, where orbital lifetimes are practically indefinite, the safety considerations are negligible. Orbits below 400 to 500 nautical miles are the ones where a safety issue is involved in case of satellite malfunction. The potential missions, the question of why reactors are being considered as a prime power candidate, reactor features, and safety considerations will be discussed

Material and component progress within ARCHER for advanced high temperature reactor

International Nuclear Information System (INIS)

Buckthorpe, D.E.; Davies, M.; Pra, F.; Bonnamy, P.; Fokkens, J.; Heijna, M.; Bout, N. de; Vreeling, A.; Bourlier, F.; Lhachemi, D.; Woayehune, A.; Dubiez-le-Goff, S.; Hahner, P.; Futterer, M.; Berka, J.; Kalivodora, J.; Pouchon, M.A.; Schmitt, R.; Homerin, P.; Marsden, B.; Mummery, P.; Mutch, G.; Ponca, D.; Buhl, P.; Hoffmann, M.; Rondet, F.; Pecherty, A.; Baurand, F.; Alenda, F.; Esch, M.; Kohlz, N.; Reed, J.; Fachinger, J.; Klower, Dr.

2014-01-01

The ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R and D) integrated project started in 2011 as part of the European Commission 7. Framework Programme (FP7) for a period of four years to perform High Temperature Reactor technology R and D in support of reactor demonstration. The project consortium encompasses conventional and Nuclear Industry, Utilities, Technical Support Organizations, Research and Development Organizations and Academia. The activities involved contribute to the Generation IV (GIF) International Forum and collaborate with related projects in the US, China, Japan, and the Republic of Korea in cooperation with IAEA and ISTC. This paper addresses the progress of the work on materials and component technologies within ARCHER over the first two years of the project. (authors)

RB research nuclear reactor, Annual report for 1982; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1981. godini

Energy Technology Data Exchange (ETDEWEB)

Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Zivkovic, B [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1982-12-15

This report includes data concerned with reactor operation and utilization, status of reactor components and equipment, refurbishment of the equipment, dosimetry and radiation protection, reactor staff, financing. It includes 7 Annexes as follows: Maintenance of reactor equipment in 1982; contents of the RB reactor safety report; review of radiation doses in the reactor building and exposure doses for the reactor staff; utilization of the RB reactor in 1982; and financial data.

The research reactors their contribution to the reactors physics

International Nuclear Information System (INIS)

Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

2000-01-01

The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01

As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

Handling of spent fuel from research reactors in Japan

International Nuclear Information System (INIS)

Kanda, K.

1997-01-01

In Japan eleven research reactors are in operation. After the 19th International Meeting on Reduced Enrichment for Research Reactors and Test Reactors (RERTR) on October 6-10, 1996, Seoul, Korea, the Five Agency Committee on Highly Enriched Uranium, which consists of Science and Technology Agency, the Ministry of Education, Science and Culture, the Ministry of Foreign Affairs, Japan Atomic Energy Research Institute (JAERI) and Kyoto University Research Reactor Institute (KURRI) met on November 7,1996, to discuss the handling of spent fuel from research reactors in Japan. Advantages and disadvantages to return spent fuel to the USA in comparison to Europe were discussed. So far, a number of spent fuel elements in JAERI and KURRI are to be returned to the US. The first shipment to the US is planned for 60 HEU elements from JMTR in 1997. The shipment from KURRI is planned to start in 1999. (author)

A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor

International Nuclear Information System (INIS)

Bretscher, M. M.

1998-01-01

The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% 235 U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm 3 and 3.8 gU/cm 3 are required to match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively

Reactor physics and reactor computations

International Nuclear Information System (INIS)

Ronen, Y.; Elias, E.

1994-01-01

Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

Mitigate Strategy of Very High Temperature Reactor Air-ingress Accident

Energy Technology Data Exchange (ETDEWEB)

Ham, Tae Kyu [KHNP CRI, Daejeon (Korea, Republic of); Arcilesi, David J.; Sun, Xiaodong; Christensen, Richard N. [The Ohio State University, Columbus (United States); Oh, Chang H.; Kim, Eung S. [Idaho National Laboratory, Idaho (United States)

2016-10-15

A critical safety event of the Very High Temperature Reactor (VHTR) is a loss-of-coolant accident (LOCA). Since a VHTR uses graphite as a core structure, if there is a break on the pressure vessel, the air in the reactor cavity could ingress into the reactor core. The worst case scenario of the accident is initiated by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. The operating pressures in the vessel and containment are about 7 and 0.1 MPa, respectively. In the VHTR, the reactor pressure vessel is located within a reactor cavity which is filled with air during normal operation. Therefore, the air-helium mixture in the cavity may ingress into the reactor pressure vessel after the depressurization process. In this paper, a commercial computational fluid dynamics (CFD) tool, FLUENT, was used to figure out air-ingress mitigation strategies in the gas-turbine modular helium reactor (GT-MHR) designed by General Atomics, Inc. After depressurization, there is almost no air in the reactor cavity; however, the air could flow back to the reactor cavity since the reactor cavity is placed in the lowest place in the reactor building. The heavier air could flow to the reactor cavity through free surface areas in the reactor building. Therefore, Argon gas injection in the reactor cavity is introduced. The injected argon would prevent the flow by pressurizing the reactor cavity initially, and eventually it prevents the flow by making the gas a heavier density than air in the reactor cavity. The gate opens when the reactor cavity is pressurized during the depressurization and it closes by gravity when the depressurization is terminated so that it can slow down the air flow to the reactor cavity.

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

The heavy water accountancy for research reactors in JAERI

International Nuclear Information System (INIS)

Yoshijima, Tetsuo; Tanaka, Sumitoshi; Nemoto, Denjirou

1998-11-01

The three research reactors have been operated by the Department of Research Reactor and used about 41 tons heavy water as coolant, moderator and reflector of research reactors. The JRR-2 is a tank type research reactor of 10MW in thermal power and its is used as moderator, coolant and reflector about 16 tons heavy water. The JRR-3M is a light water cooled and moderated pool type research reactor with a thermal power of 20MW and its is used as reflector about 7.3 tons heavy water. In the JRR-4, which is a light water cooled swimming pool type research reactor with the maximum thermal power of 3.5MW, about 1 ton heavy water is used to supply fully thermalized neutrons with a neutron beam experiment of facility. The heavy water was imported from U.S.A., CANADA and Norway. Parts of heavy water is internationally controlled materials, therefore management of heavy water is necessary for materials accountancy. This report described the change of heavy water inventories in each research reactors, law and regulations for accounting of heavy water in JAERI. (author)

Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

International Nuclear Information System (INIS)

Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

2008-01-01

TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2.7

United States Domestic Research Reactor Infrastructure - TRIGA Reactor Fuel Support

International Nuclear Information System (INIS)

Morrell, Douglas

2008-01-01

The purpose of the United State Domestic Research Reactor Infrastructure Program is to provide fresh nuclear reactor fuel to United States universities at no, or low, cost to the university. The title of the fuel remains with the United States government and when universities are finished with the fuel, the fuel is returned to the United States government. The program is funded by the United States Department of Energy - Nuclear Energy division, managed by Department of Energy - Idaho Field Office, and contracted to the Idaho National Laboratory's Management and Operations Contractor - Battelle Energy Alliance. Program has been at Idaho since 1977 and INL subcontracts with 26 United States domestic reactor facilities (13 TRIGA facilities, 9 plate fuel facilities, 2 AGN facilities, 1 Pulstar fuel facility, 1 Critical facility). University has not shipped fuel since 1968 and as such, we have no present procedures for shipping spent fuel. In addition: floor loading rate is unknown, many interferences must be removed to allow direct access to the reactor tank, floor space in the reactor cell is very limited, pavement ends inside our fence; some of the surface is not finished. The whole approach is narrow, curving and downhill. A truck large enough to transport the cask cannot pull into the lot and then back out (nearly impossible / refused by drivers); a large capacity (100 ton), long boom crane would have to be used due to loading dock obstructions. Access to the entrance door is on a sidewalk. The campus uses it as a road for construction equipment, deliveries and security response. Large trees are on both sides of sidewalk. Spent fuel shipments have never been done, no procedures approved or in place, no approved casks, no accident or safety analysis for spent fuel loading. Any cask assembly used in this facility will have to be removed from one crane, moved on the floor and then attached to another crane to get from the staging area to the reactor room. Reactor

Requirements for light water reactors

International Nuclear Information System (INIS)

Hedin, F.

2009-01-01

The EUR (European Utilities Requirements) is an organization founded in 1991 whose aim was to write down the European specifications and requirements for the future reactors of third generation. EUR gathers most of the nuclear power producers of Europe. The EUR document has been built on the large and varied experience of EUR members and can be used to elaborate invitations to tender for nuclear projects. 4000 requirements only for the nuclear part of the plant are listed, among which we have: -) the probability of core meltdown for a reactor must be less than 10 -6 per year, -) the service life of every component that is not replaceable must be 60 years, -) the capacity of the spent fuel pool must be sufficient to store 10-15 years of production without clearing out. The EUR document is both open and complete: every topic has been considered, it does not favor any type of reactor but can ban any technology that is too risky or has an unfavourable feedback experience. The assessment of the conformity with the EUR document of 7 reactor projects (BWR 90/, EPR, EP1000, SWR1000, ABWR, AP1000 and VVER-AES-92) has already be made. (A.C.)

Reactor

International Nuclear Information System (INIS)

Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

1979-01-01

Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

Reactor Physics Training

International Nuclear Information System (INIS)

Baeten, P.

2007-01-01

University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

FBR type reactor

International Nuclear Information System (INIS)

Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

1994-01-01

In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

Reactor container

International Nuclear Information System (INIS)

Kato, Masami; Nishio, Masahide.

1987-01-01

Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

Review of the United Kingdom fast reactor programme - March 1986

International Nuclear Information System (INIS)

Bramman, J.I.; John, C.T.; Wheeler, R.C.

1986-01-01

The UK programme in the field of fast reactors has continued successfully towards the following main objectives, details of which are contained in subsequent sections of this report: (2) progress with the prototype fast reactor (PFR) which achieved its design power on 4 March 1985; (3) nuclear fuel reprocessing; (4) commercial design studies; (5) structural integrity of LMFBR during its lifetime; (6) R and D work on components of LMFBR; (7) materials study; (8) sodium chemistry; (9) reactor core and fuel design philosophy; (10) safety problems; (11) plant performance studies

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

Detection of pressure tube leaks relying on moisture beetles only

International Nuclear Information System (INIS)

Kenchington, J.M.; Choi, A.; Jin, Y.

2004-01-01

A major decision was made for Pickering NGS A Annulus Gas System (ACS) that detection of a pressure tube (PT) leak should be achieved by using only moisture beetles and that dew point monitors would provide 'early warning' without status to shut down the reactor. Experience with Unit 3 has shown that dew point monitoring of pressure tube leaks was particularly subject to gas leaks and surface adsorption effects. Unit 4 was the first one to be converted during the full scale pressure tube replacement programme. Because of the fundamental change in design philosophy, moisture injection tests were carried out during commissioning to demonstrate that performance matched design. In particular it was necessary to show that leak before break (LBB) would be achieved if a leak occurred in the limiting string. Units 1 and 3 have since been converted. No decision has been taken to convert Pickering B units as gas leaks are small and no significant adsorption effects are anticipated. Hence dew point monitoring will not be impaired. (author)

The Integration Of Process Heat Applications To High Temperature Gas Reactors

International Nuclear Information System (INIS)

McKellar, Michael G.

2011-01-01

A high temperature gas reactor, HTGR, can produce industrial process steam, high-temperature heat-transfer gases, and/or electricity. In conventional industrial processes, these products are generated by the combustion of fossil fuels such as coal and natural gas, resulting in significant emissions of greenhouse gases such as carbon dioxide. Heat or electricity produced in an HTGR could be used to supply process heat or electricity to conventional processes without generating any greenhouse gases. Process heat from a reactor needs to be transported by a gas to the industrial process. Two such gases were considered in this study: helium and steam. For this analysis, it was assumed that steam was delivered at 17 MPa and 540 C and helium was delivered at 7 MPa and at a variety of temperatures. The temperature of the gas returning from the industrial process and going to the HTGR must be within certain temperature ranges to maintain the correct reactor inlet temperature for a particular reactor outlet temperature. The returning gas may be below the reactor inlet temperature, ROT, but not above. The optimal return temperature produces the maximum process heat gas flow rate. For steam, the delivered pressure sets an optimal reactor outlet temperature based on the condensation temperature of the steam. ROTs greater than 769.7 C produce no additional advantage for the production of steam.

Socioeconomic consequences of nuclear reactor accidents

International Nuclear Information System (INIS)

Tawil, J.J.; Callaway, J.W.; Coles, B.L.; Cronin, F.J.; Currie, J.W.; Imhoff, K.L.; Lewis, P.M.; Nesse, R.J.; Strenge, D.L.

1984-06-01

This report identifies and characterizes the off-site socioeconomic consequences that would likely result from a severe radiological accident at a nuclear power plant. The types of impacts that are addressed include economic impacts, health impacts, social/psychological impacts and institutional impacts. These impacts are identified for each of several phases of a reactor accident - from the warning phase through the post-resettlement phase. The relative importance of the impact during each accident phase and the degree to which the impact can be predicted are indicated. The report also examines the methods that are currently used for assessing nuclear reactor accidents, including development of accident scenarios and the estimating of socioeconomic accident consequences with various models. Finally, a critical evaluation is made regarding the use of impact analyses in estimating the contribution of socioeconomic consequences to nuclear accident reactor accident risk. 116 references, 7 figures, 15 tables

Low power reactor for remote applications

International Nuclear Information System (INIS)

Meier, K.L.; Palmer, R.G.; Kirchner, W.L.

1985-01-01

A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long-term, virtually maintenance free, operation of this reactor for remote applications. 10 refs., 7 figs., 3 tabs

Water Chemistry Control in Reducing Corrosion and Radiation Exposure at PWR Reactor

International Nuclear Information System (INIS)

Febrianto

2006-01-01

Water chemistry control plays an important role in relation to plant availability, reliability and occupational radiation exposures. Radiation exposures of nuclear plant workers are determined by the radiation rate dose and by the amount maintenance and repair work time Water chemistry has always been, from beginning of operation of power Pressurized Water Reactor, an important factor in determining the integrity of reactor components, fuel cladding integrity and minimize out of core radiation exposures. For primary system, the parameters to control the quality of water chemistry have been subject to change in time. Reactor water coolant pH need to be optimally controlled and be operated in range pH 6.9 to 7.4. At pH lower than 6.9, cause increasing the radiation exposure level and increasing coolant water pH higher than 7.4 will decrease radiation exposure level but increasing risk to fuel cladding and steam generator tube. Since beginning 90 decade, PWR water coolant pH tend to be operated at pH 7.4. This paper will discuss concerning water chemistry development in reducing corrosion and radiation exposure dose in PWR reactor. (author)

Office for Analysis and Evaluation of Operational Data 1992 annual report: Power reactors. Volume 7, No. 1

Energy Technology Data Exchange (ETDEWEB)

None

1993-07-01

The annual report of the US Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during 1992. The report is published in two separate parts. NUREG-1272, Vol. 7, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about the trends of some key performance, measures. The report also includes the principal findings and issues identified in AEOD studies over the past year, and summarizes information from such sources as licensee event report% diagnostic evaluations, and reports to the NRC`s Operations Center. The reports contain a discussion of the Incident Investigation Team program and summarize the Incident Investigation Team and Augmented Inspection Team reports for that group of licensees. NUREG-1272, Vol. 7, No. 2, covers nonreactors and presents a review of the events and concerns during 1992 associated with the use of licensed material in nonreactor applications, such as personnel overexposures and medical misadministrations. Each volume contains a list of the AEOD reports issued for 1984--1992.

Nuclear Reactor RA Safety Report, Format and Contents

International Nuclear Information System (INIS)

1986-11-01

This is a new complete version of the safety report of nuclear reactor RA is made according to the recommendations of the IAEA. Report includes all the relevant data needed for evaluation of safe operation of this nuclear facility. Each of seven volumes of this report cover separate topics as follows: (1) introduction; (2) Site characteristics; (3) description of the reactor building and installations; (4) description of the reactor; (5) description of the coolant system; (6) description of the regulation and safety instrumentation; (7) description of the power supply system; (8) description of the auxiliary systems; (9) radiation protection issues; (10) radioactive waste management (11) reactor operation; (12) accident analysis during previous operation; (13) analysis of possible accident causes; (14) safety analysis and preventive actions: (15) analysis of significant accidents; (16) analysis of maximum possible accident; (17) environmental impact analysis in case of accident [sr

CANDU development

International Nuclear Information System (INIS)

Brooks, G.L.

1981-06-01

Evolution of the 950 MW(e) CANDU reactor is summarized. The design was specifically aimed at the export market. Factors considered in the design were that 900-1000 MW is the maximum practical size for most countries; many countries have warmer condenser cooling water than Canada; the plant may be located on coastal sites; seismic requirements may be more stringent; and the requirements of international, as well as Canadian, standards must be satisfied. These considerations resulted in a 600-channel reactor capable of accepting condenser cooling water at 32 0 C. To satisfy the requirement for a proven design, the 950 MW CANDU draws upon the basic features of the Bruce and Pickering plants which have demonstrated high capacity factors

Effects of variable-row-spacing harvesting picker platform scraping plates on cotton fiber quality and quantity

Directory of Open Access Journals (Sweden)

Cíntia Michele de Campos Baraviera

2017-06-01

Full Text Available There have been increasing demands for high-quality cotton fibers that meet the textile industry quality standards. Concurrently, there have been efforts to reduce contaminants during harvesting to reduce harvesting costs. The goal of this research was to evaluate the efficiency of the picker platform with Variable-Row-Spacing (VRS for harvesting cotton in narrow rows, over two harvest seasons in two regions within the state of Mato Grosso, Brazil. In this study, how the presence vs. absence of scraping plates and variations in travel speed was related to quantifiable levels of impurities the harvested fibers was examined. The research was divided into three experiments (Exp. I, II, and III, using cotton varieties FM 975 WS, IMA 5672 B2 RF, and IMA 5675 B2 RF, with row spacing of 0.45 m. The experimental design was randomized blocks, in a 2 ? 3 factorial design, using the presence/absence of the plate and three speeds (0.61, 1.0, and 1.42 m·s-¹, with seven repetitions, totaling 42 experimental plots. The plot size was 108 m² (3.6 ? 30 m. The data were analyzed using the F test in ANOVA and the post-hoc Tukey test (p < 0.05. The results showed that scraping plates increased the number of stems and cones, and reduced the harvest efficiency of cotton planted in narrow rows in the region of Sorriso-MT during the 2013/2014 harvest. For the 2014/2015 harvest, the highest speed and the presence of the scraping plates increased the number of cones in the cotton samples. In the experiment conducted in Serra da Petrovina, the removal of the scraping plates decreased the amount of cones in the harvested cotton.

Fabrication of magnetic hydroxypropyl cellulose-g-poly(acrylic acid) porous spheres via Pickering high internal phase emulsion for removal of Cu(2+) and Cd(2.).

Science.gov (United States)

Zhu, Yongfeng; Zheng, Yian; Zong, Li; Wang, Feng; Wang, Aiqin

2016-09-20

A series of magnetic hydroxypropyl cellulose-g-poly(acrylic acid) porous spheres were prepared via O/W Pickering high internal phase emulsions (HIPEs) integrated precipitation polymerization. The structure and composition of modified Fe3O4 and porous structures were characterized by TEM, XRD, TGA and SEM. The results indicated that the silanized Fe3O4 can influence greatly the pore structure of magnetic porous sphere in addition to non-negligible impacts of the proportion of mixed solvent and co-surfactant. The adsorption experiment demonstrated that the adsorption equilibrium can be reached within 40min and the maximal adsorption capacity was 300.00mg/g for Cd(2+) and 242.72mg/g for Cu(2+), suggesting its fast adsorption kinetics and high adsorption capacity. After five adsorption-desorption cycles, no significant changes in the adsorption capacity were observed, suggesting its excellent reusability. The magnetic porous sphere can be easily separated from the solution and then find its potential as a recyclable material for highly efficient removal of heavy metals. Copyright © 2016 Elsevier Ltd. All rights reserved.

The rehabilitation/upgrading of Philippine Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Renato, T Banaga [Philippines Nuclear Research Inst., Quezon (Philippines)

1998-10-01

The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E{sub 1}-U-Z{sub 1}-H{sub 1.6} TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

The rehabilitation/upgrading of Philippine Research Reactor

International Nuclear Information System (INIS)

Renato T, Banaga

1998-01-01

The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E 1 -U-Z 1 -H 1.6 TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

HTR-2002: Proceedings of the conference on high temperature reactors

International Nuclear Information System (INIS)

2002-01-01

High temperature reactors are considered as future inherently safe and efficient energy sources. The presentations covered all the relevant aspects of the existing HTGRs and/or helium cooled pebble bed reactors. They were sorted into 7 sessions: HTR Projects and Programmes; Fuel and Fuel Cycle; Physics and Neutronics; Thermohydraulic Calculation; Engineering, Design and Applications; Materials and Components; Safety and Licensing

Nuclear reactors. Introduction

International Nuclear Information System (INIS)

Boiron, P.

1997-01-01

This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

Burnout experiments with 6 x 6, 8 x 8 and 7 x 7 rod bundle test sections using freon as model fluid

International Nuclear Information System (INIS)

Fulfs, H.; Katsaounis, A.; Minden, C.v.

1976-01-01

This paper reports on burnout experiments at staedy state condition using Freon12 as model fluid. The experiments were carried out with three test sections with 6 x 6, 8 x 8 and 7 x 7 rod bundles. The axial flux distribution of the rods is either constant or reactor like. The transformed measured points using STEVENS and BOURE scaling factors to equivalent water conditions respectively, were compared to the burnout correlation W3 using the reactor layout program DYNAMIT. The DYNAMIT code is a thermohydraulic lay-out reactor program without consideration of mixing flow between the subchannels. (orig.) [de

Nuclear reactor, reactor core thereof, and device for constituting the reactor

International Nuclear Information System (INIS)

Takiyama, Masashi.

1994-01-01

A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

Cavity temperature and flow characteristics in a gas-core test reactor

Science.gov (United States)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

Slurry reactors

Energy Technology Data Exchange (ETDEWEB)

Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

1979-08-01

Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

Optimal reactor strategy for commercializing fast breeder reactors

International Nuclear Information System (INIS)

Yamaji, Kenji; Nagano, Koji

1988-01-01

In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)